Evaluation of actinide partitioning and transmutation in light-water reactors

International Nuclear Information System (INIS)

Collins, Emory D.; Renier, John-Paul

2004-01-01

Advanced Fuel Cycle Initiative (AFCI) studies were made to evaluate the feasibility of multicycle transmutation of plutonium and the minor actinides (MAs) in light-water reactors (LWRs). Results showed that significant repository benefits, cost reductions, proliferation resistance, and effective use of facilities can be obtained. Key advantages are shown to be made possible by processing 30-year-decayed spent fuel rather than the more traditional 5-year-decayed fuel. (authors)

New lineup of light water reactors

International Nuclear Information System (INIS)

Okamura, Kiyoshi; Oshima, Koichiro; Kitsukawa, Keisuke

2007-01-01

Toshiba is promoting technical studies for upcoming nuclear power plants based on its large accumulation of experience in boiling water reactor (BWR) design, manufacturing, construction, and maintenance. Our goal is to achieve higher reliability, lower life-cycle costs, and better competitiveness for nuclear power plants compared with other energy sources. In addition, we are developing a new light water reactor (LWR) lineup featuring the safest and most economical LWRs in the world as next-generation reactors almost at new construction and replacement in the Japanese and international markets expected to start from the 2020s. We are committed not only to developing BWRs with the world's highest performance but also to participating in the pressurized water reactor (PWR) market, taking advantage of the synergistic effect of both Toshiba's and Westinghouse's experience. (author)

Advanced Light Water Reactor Program: Program management and staff review methodology

International Nuclear Information System (INIS)

Moran, D.H.

1986-12-01

This report summarizes the NRC/EPRI coordinated effort to develop design requirements for a standardized advanced light water reactor (ALWR) and the procedures for screening and applying new generic safety issues to this program. The end-product will be an NRC-approved ALWR Requirements Document for use by the nuclear industry in generating designs of LWRs to be constructed for operation in the 1990s and beyond

Application of Dr. Mainte, integrated simulator of maintenance optimization, to LWRs

International Nuclear Information System (INIS)

Isobe, Yoshihiro; Sagisaka, Mitsuyuki; Etoh, Junji; Matsunaga, Takashi; Kosaka, Toru; Matsumoto, Satoshi; Yoshimura, Shinobu

2015-01-01

Dr. Mainte, an integrated simulator for maintenance optimization of LWRs (Light Water Reactors) is based on PFM (Probabilistic Fracture Mechanics) analyses. The concept of the simulator is to provide a decision-making system to optimize maintenance activities for typical components and piping systems in nuclear power plants totally and quantitatively in terms of safety, availability, economic rationality, environmental impact and social acceptance. For the further improvement of the safety and availability of nuclear power plants, the effect of human error and its reduction on the optimization of maintenance activities have been studied. In addition, an approach of reducing human error is proposed. (author)

Environmentally assisted cracking in Light Water Reactors

International Nuclear Information System (INIS)

Chung, H.M.; Chopra, O.K.; Ruther, W.E.; Kassner, T.F.; Michaud, W.F.; Park, J.Y.; Sanecki, J.E.; Shack, W.J.

1993-09-01

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1992 to March 1993. Fatigue and EAC of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (1) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels. (2) EAC of cast stainless steels (SSs), (3) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence, and (4) EAC of low-alloy steels. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions and chromium-nickel-plated A533-Gr B steel in simulated boiling-water reactor (BWR) water at 289 degrees C. The data were compared with predictions based on crack growth correlations for ferritic steels in oxygenated water and correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy

Passive and inherent safety technologies for light-water nuclear reactors

International Nuclear Information System (INIS)

Forsberg, C.W.

1990-07-01

Passive/inherent safety implies a technical revolution in our approach to nuclear power safety. This direction is discussed herein for light-water reactors (LWRs) -- the predominant type of power reactor used in the world today. At Oak Ridge National Laboratory (ORNL) the approach to the development of passive/inherent safety for LWRs consists of four steps: identify and quantify safety requirements and goals; identify and quantify the technical functional requirements needed for safety; identify, invent, develop, and quantify technical options that meet both of the above requirements; and integrate safety systems into designs of economic and reliable nuclear power plants. Significant progress has been achieved in the first three steps of this program. The last step involves primarily the reactor vendors. These activities, as well as related activities worldwide, are described here. 27 refs., 7 tabs

Materials Inventory Database for the Light Water Reactor Sustainability Program

Energy Technology Data Exchange (ETDEWEB)

Kazi Ahmed; Shannon M. Bragg-Sitton

2013-08-01

Scientific research involves the purchasing, processing, characterization, and fabrication of many sample materials. The history of such materials can become complicated over their lifetime – materials might be cut into pieces or moved to various storage locations, for example. A database with built-in functions to track these kinds of processes facilitates well-organized research. The Material Inventory Database Accounting System (MIDAS) is an easy-to-use tracking and reference system for such items. The Light Water Reactor Sustainability Program (LWRS), which seeks to advance the long-term reliability and productivity of existing nuclear reactors in the United States through multiple research pathways, proposed MIDAS as an efficient way to organize and track all items used in its research. The database software ensures traceability of all items used in research using built-in functions which can emulate actions on tracked items – fabrication, processing, splitting, and more – by performing operations on the data. MIDAS can recover and display the complete history of any item as a simple report. To ensure the database functions suitably for the organization of research, it was developed alongside a specific experiment to test accident tolerant nuclear fuel cladding under the LWRS Advanced Light Water Reactor Nuclear Fuels Pathway. MIDAS kept track of materials used in this experiment from receipt at the laboratory through all processes, test conduct and, ultimately, post-test analysis. By the end of this process, the database proved to be right tool for this program. The database software will help LWRS more efficiently conduct research experiments, from simple characterization tests to in-reactor experiments. Furthermore, MIDAS is a universal tool that any other research team could use to organize their material inventory.

Status and subjects of thermal-hydraulic analysis for next-generation LWRs

International Nuclear Information System (INIS)

2000-03-01

The status and subjects on thermal-hydraulic analysis for next-generation light water reactors (LWRs) with passive safety systems were surveyed through about 5 years until March 1999 by subcommittee on improvement of reactor thermal-hydraulic analysis codes under the nuclear code committee in Japan Atomic Energy Research Institute. Based on the survey results and discussion, the status and subjects on system analysis for various types of proposed reactor were summarized in 1998 and those on multidimensional two-phase flow analysis were also reviewed, since the multidimensional analysis was recognized as one of the most important subjects through the investigation on system analysis. In this report, the status and subjects for the following were summarized from the survey results and discussion in 1998 and 1999; (1) BWR neutronic/thermal-hydraulic coupled analysis, (2) Evaluation of passive safety system performance and (3) Gas-liquid two-phase flow analysis. The contents in this report are the forefront of thermal-hydraulic analysis for LWRs including test results from several large-scale facilities. We expect that the contents can offer a guideline to improve reactor thermal-hydraulic analysis codes in future. (author)

Assessment of nonbackfittable concepts for improving uranium utilization in LWRs

International Nuclear Information System (INIS)

Newman, D.F.; Goldsmith, S.; Fleischman, R.M.

1980-01-01

Recent efforts to improve uranium utilization in light water reactors (LWRs) have involved backfittable changes to fuel or operations. The Advanced Reactor Design Study sponsored by the US Department of Energy identified and evaluated nonbackfittable LWR concepts to provide a basis for selecting and demonstrating specific improvements that have good implementation potential. Because the application of nonbackfittable concepts necessitates modifications to contemporary reactor designs, it was apparent that the most qualified organizations to assess implementation potential would be LWR designers/vendors. Accordingly, Babcock and Wilcox, Combustion Engineering, and General Electric were the principal participants in selecting, assessing, and evaluating the nonbackfittable concepts included in this study. The results of the industrial assessments of nonbackfittable LWR concepts are shown

Uranium utilization of light water cooled reactors and fast breeders

International Nuclear Information System (INIS)

Stojadinovic, Timm

1991-08-01

The better uranium utilization of fast breeder reactors as compared with water cooled reactors is one argument in favour of the breeder introduction. This report tries to quantify this difference. It gives a generally valid formalism for the uranium utilization as a function of the fuel burnup, the conversion rate, fuel cycle losses and the fuel enrichment. On the basis of realistic assumptions, the ratio between the utilizations of breeder reactors to that of light water cooled reactors (LWR) amounts to 180 for the open LWR cycle and 100 in case of plutonium recycling in LWRs

Verification and Validation Strategy for LWRS Tools

Energy Technology Data Exchange (ETDEWEB)

Carl M. Stoots; Richard R. Schultz; Hans D. Gougar; Thomas K Larson; Michael Corradini; Laura Swiler; David Pointer; Jess Gehin

2012-09-01

One intension of the Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) program is to create advanced computational tools for safety assessment that enable more accurate representation of a nuclear power plant safety margin. These tools are to be used to study the unique issues posed by lifetime extension and relicensing of the existing operating fleet of nuclear power plants well beyond their first license extension period. The extent to which new computational models / codes such as RELAP-7 can be used for reactor licensing / relicensing activities depends mainly upon the thoroughness with which they have been verified and validated (V&V). This document outlines the LWRS program strategy by which RELAP-7 code V&V planning is to be accomplished. From the perspective of developing and applying thermal-hydraulic and reactivity-specific models to reactor systems, the US Nuclear Regulatory Commission (NRC) Regulatory Guide 1.203 gives key guidance to numeric model developers and those tasked with the validation of numeric models. By creating Regulatory Guide 1.203 the NRC defined a framework for development, assessment, and approval of transient and accident analysis methods. As a result, this methodology is very relevant and is recommended as the path forward for RELAP-7 V&V. However, the unique issues posed by lifetime extension will require considerations in addition to those addressed in Regulatory Guide 1.203. Some of these include prioritization of which plants / designs should be studied first, coupling modern supporting experiments to the stringent needs of new high fidelity models / codes, and scaling of aging effects.

Anticipated transients without scram for light water reactors: implications for liquid metal fast breeder reactors

International Nuclear Information System (INIS)

Kastenberg, W.E.; Solomon, K.A.

1979-07-01

In the design of light water reactors (LWRs), protection against anticipated transients (e.g., loss of normal electric power and control rod withdrawal) is provided by a highly reliable scram, or shutdown system. If this system should become inoperable, however, the transient could lead to a core meltdown. The Nuclar Regulatory Commission (NRC) has proposed, in NUREG-0460 [1], new requirements (or acceptance criteria) for anticipated transients without scram (ATWS) events and the manner in which they could be considered in the design and safety evaluation of LWRs. This note assesses the potential impact of the proposed LWR-ATWS criteria on the liquid metal fast breeder reactor (LMFBR) safety program as represented by the Clinch River Breeder Reactor Plant

Flow-induced vibration for light-water reactors. Progress report, April 1978-December 1979

International Nuclear Information System (INIS)

Schardt, J.F.

1980-03-01

Flow-Induced vibration for Light Water Reactors (FIV for LWRs) is a four-year program designed to improve the FIV performance of light water reactors through the development of design criteria, analytical models for predicting behavior of components, general scaling laws to improve the accuracy of reduced-scale tests, and the identification of high FIV risk areas. The program commenced December 1, 1976, but was suspended on September 30, 1978, due to a shift in Department of Energy (DOE) priorities away from LWR productivity/availability. It was reinitiated as of August 1, 1979. This progress report summarizes the accomplishments achieved during the period from April 1978 to December 1979

Nondestructive examination (NDE) Reliability for Inservice Inspection of Light Water Reactors

International Nuclear Information System (INIS)

Doctor, S.R.; Good, M.S.; Heasler, P.G.; Hockey, R.L.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Vo, T.V.

1992-07-01

The Evaluation and Improvement of NDE reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to the Regulatory and ASME Code requirements, based on material properties, service conditions, and NDE uncertainties

Thermophysical properties database of materials for light water reactors and heavy water reactors. Final report of a coordinated research project 1999-2005

International Nuclear Information System (INIS)

2006-06-01

The IAEA Coordinated Research Project (CRP) on the Establishment of a Thermo-physical Properties Database for Light Water Reactors (LWRs) and Heavy Water Reactors (HWRs) started in 1999. It was included in the IAEA's Nuclear Power Programme following endorsement in 1997 by the IAEA's Technical Working Groups on Advanced Technologies for LWRs and HWRs (the TWG-LWR and the TWG-HWR). Furthermore, the TWG on Fuel Performance and Technology (TWG-FPT) also expressed its support. This CRP was conducted as a joint task within the IAEA's project on technology development for LWRs and HWRs in its nuclear power programme. Improving the technology for nuclear reactors through better computer codes and more accurate materials property data can contribute to improved economics of future plants by helping to remove the need for large design margins, which are currently used to account for limitations of data and methods. Accurate representations of thermo-physical properties under relevant temperature and neutron fluence conditions are necessary for evaluating reactor performance under normal operation and accident conditions. The objective of this CRP was to collect and systematize a thermo-physical properties database for light and heavy water reactor materials under normal operating, transient and accident conditions and to foster the exchange of non-proprietary information on thermo-physical properties of LWR and HWR materials. An internationally available, peer reviewed database of properties at normal and severe accident conditions has been established on the Internet. This report is intended to serve as a useful source of information on thermo-physical properties data for water cooled reactor analyses. The properties data have been initially stored in the THERSYST data system at the University of Stuttgart, Germany, which was subsequently developed into an internationally available Internet database named THERPRO at Hanyang University, Republic of Korea

Resonance self-shielding method using resonance interference factor library for practical lattice physics computations of LWRs

International Nuclear Information System (INIS)

Choi, Sooyoung; Khassenov, Azamat; Lee, Deokjung

2016-01-01

This paper presents a new method of resonance interference effect treatment using resonance interference factor for high fidelity analysis of light water reactors (LWRs). Although there have been significant improvements in the lattice physics calculations over the several decades, there exist still relatively large errors in the resonance interference treatment, in the order of ∼300 pcm in the reactivity prediction of LWRs. In the newly developed method, the impact of resonance interference to the multi-group cross-sections has been quantified and tabulated in a library which can be used in lattice physics calculation as adjustment factors of multi-group cross-sections. The verification of the new method has been performed with Mosteller benchmark, UO_2 and MOX pin-cell depletion problems, and a 17×17 fuel assembly loaded with gadolinia burnable poison, and significant improvements were demonstrated in the accuracy of reactivity and pin power predictions, with reactivity errors down to the order of ∼100 pcm. (author)

Thorium Fuel Performance in a Tight-Pitch Light Water Reactor Lattice

International Nuclear Information System (INIS)

Kim, Taek Kyum; Downar, Thomas J.

2002-01-01

Research on the utilization of thorium-based fuels in the intermediate neutron spectrum of a tight-pitch light water reactor (LWR) lattice is reported. The analysis was performed using the Studsvik/Scandpower lattice physics code HELIOS. The results show that thorium-based fuels in the intermediate spectrum of tight-pitch LWRs have considerable advantages in terms of conversion ratio, reactivity control, nonproliferation characteristics, and a reduced production of long-lived radiotoxic wastes. Because of the high conversion ratio of thorium-based fuels in intermediate spectrum reactors, the total fissile inventory required to achieve a given fuel burnup is only 11 to 17% higher than that of 238 U fertile fuels. However, unlike 238 U fertile fuels, the void reactivity coefficient with thorium-based fuels is negative in an intermediate spectrum reactor. This provides motivation for replacing 238 U with 232 Th in advanced high-conversion intermediate spectrum LWRs, such as the reduced-moderator reactor or the supercritical reactor

Studies of severe accidents in light-water reactors

International Nuclear Information System (INIS)

1987-01-01

From 10 to 12 November 1986 some 80 delegates met under the auspices of the CEC working group on the safety of light-water reactors. The participants from EC Member States were joined by colleagues from Sweden, Finland and the USA and met to discuss the subject of severe accidents in LWRs. Although this seminar had been planned well before Chernobyl, the ''severe-accident-that-really-happened'' made its mark on the seminar. The four main seminar topics were: (i) high source-term accident sequences identified in PSAs, (ii) containment performance, (iii) mitigation of core melt consequences, (iv) severe accident management in LWRs. In addition to the final panel discussion there was also a separate panel discussion on lessons learned from the Chernobyl accident. These proceedings include the papers presented during the seminar and they are arranged following the seminar programme outline. The presentations and discussions of the two panels are not included in the proceedings. The general conclusions and directions following from these two panels were, however, considered in a seminar review paper which was published in the March 1987 issue of Nuclear Engineering International

A review of multiple recycle of plutonium in LWRs

International Nuclear Information System (INIS)

Hesketh, K.

1997-01-01

This paper presents a summary of the current status of the nuclear physics of multiple recycle in Light Water Reactors (LWRs). The paper draws attention to the principal nuclear physics issues of relevance and highlights where present knowledge is inadequate. This paper reviews the existing literature critically and provides a clearer picture of the assumptions underlying the past work and will highlight areas where past work has been deficient. From this comes a concise analysis of the technicalities of multiple MOX recycle and the areas of uncertainty which remain. The paper concludes that at present the question remains one for which there is as yet no satisfactory answer, largely because the basic nuclear data and lattice codes essential to define the technical limits are not yet sufficiently developed and validated. 6 refs, 2 figs, 3 tabs

Status of advanced technology and design for water cooled reactors: Light water reactors

International Nuclear Information System (INIS)

1988-10-01

Water reactors represent a high level of performance and safety. They are mature technology and they will undoubtedly continue to be the main stream of nuclear power. There are substantial technological development programmes in Member States for further improving the technology and for the development of new concepts in water reactors. Therefore the establishment of an international forum for the exchange of information and stimulation of international co-operation in this field has emerged. In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors has been undertaken to document the major current activities and different trends of technological improvements and developments for future water reactors. Part I of the report dealing with LWRs has now been prepared and is based mainly on submissions from Member States. It is hoped that this part of the report, containing the status of advanced light water reactor design and technology of the year 1987 and early 1988 will be useful for disseminating information to Agency Member States and for stimulating international cooperation in this subject area. 93 refs, figs and tabs

Identification and characterization of passive safety system and inherent safety feature building blocks for advanced light-water reactors

International Nuclear Information System (INIS)

Forsberg, C.W.

1989-01-01

Oak Ridge National Laboratory (ORNL) is investigating passive and inherent safety options for Advanced Light-Water Reactors (ALWRs). A major activity in 1989 includes identification and characterization of passive safety system and inherent safety feature building blocks, both existing and proposed, for ALWRs. Preliminary results of this work are reported herein. This activity is part of a larger effort by the US Department of Energy, reactor vendors, utilities, and others in the United States to develop improved LWRs. The Advanced Boiling Water Reactor (ABWR) program and the Advanced Pressurized Water Reactor (APWR) program have as goals improved, commercially available LWRs in the early 1990s. The Advanced Simplified Boiling Water Reactor (ASBWR) program and the AP-600 program are developing more advanced reactors with increased use of passive safety systems. It is planned that these reactors will become commercially available in the mid 1990s. The ORNL program is an exploratory research program for LWRs beyond the year 2000. Desired long-term goals for such reactors include: (1) use of only passive and inherent safety, (2) foolproof against operator errors, (3) malevolence resistance against internal sabotage and external assault and (4) walkaway safety. The acronym ''PRIME'' [Passive safety, Resilient operation, Inherent safety, Malevolence resistance, and Extended (walkaway) safety] is used to summarize these desired characteristics. Existing passive and inherent safety options are discussed in this document

Key physical parameters and temperature reactivity coefficients of the deep burn modular helium reactor fueled with LWRs waste

Energy Technology Data Exchange (ETDEWEB)

Talamo, Alberto E-mail: alby@neutron.kth.se; Gudowski, Waclaw E-mail: wacek@neutron.kth.se; Cetnar, Jerzy E-mail: jerzy@neutron.kth.se; Venneri, Francesco E-mail: venneri@lanl.gov

2004-11-01

We investigated some important neutronic features of the deep burn modular helium reactor (DB-MHR) using the MCNP/MCB codes. Our attention was focused on the neutron flux and its spectrum, capture to fission ratio of {sup 239}Pu and the temperature coefficient of fuel and moderator. The DB-MHR is a graphite-moderated helium-cooled reactor proposed by General Atomic to address the need for a fast and efficient incineration of plutonium for non-proliferation purposes as well as the management of light water reactors (LWRs) waste. In fact, recent studies have shown that the use of the DB-MHR coupled to ordinary LWRs would keep constant the world inventory of plutonium for a reactor fleet producing 400 TW{sub e}/y. In the present studies, the DB-MHR is loaded with Np-Pu driver fuel (DF) with an isotopic composition corresponding to LWRs spent fuel waste. DF uses fissile isotopes (e.g. {sup 239}Pu and {sup 241}Pu), previously generated in the LWRs, and maintains criticality conditions in the DB-MHR. After an irradiation of three years, the spent DF is reprocessed and its remaining actinides are manufactured into fresh transmutation fuel (TF). TF mainly contains non-fissile actinides which undergo neutron capture and transmutation during the subsequent three-year irradiation in the DB-MHR. At the same time, TF provides control and negative reactivity feedback to the reactor. After extraction of the spent TF, irradiated for three years, over 94% of {sup 239}Pu and 53% of all actinides coming from LWRs waste will have been destroyed in the DB-MHR. In this paper we look at the operation conditions at equilibrium for the DB-MHR and evaluate fluxes and reactivity responses using state of the art 3-D Monte Carlo simulations.

Current status of materials development of nuclear fuel cladding tubes for light water reactors

International Nuclear Information System (INIS)

Duan, Zhengang; Yang, Huilong; Satoh, Yuhki; Murakami, Kenta; Kano, Sho; Zhao, Zishou; Shen, Jingjie; Abe, Hiroaki

2017-01-01

Zirconium-based (Zr-based) alloys have been widely used as materials for the key components in light water reactors (LWRs), such as fuel claddings which suffer from waterside corrosion, hydrogen uptakes and strength loss at elevated temperature, especially during accident scenarios like the lost-of-coolant accident (LOCA). For the purpose of providing a safer, nuclear leakage resistant and economically viable LWRs, three general approaches have been proposed so far to develop the accident tolerant fuel (ATF) claddings: optimization of metallurgical composition and processing of Zr-based alloys, coatings on existing Zr-based alloys and replacement of current Zr-based alloys. In this manuscript, an attempt has been made to systematically present the historic development of Zr-based cladding, including the impacts of alloying elements on the material properties. Subsequently, the research investigations on coating layer on the surface of Zr-based claddings, mainly referring coating materials and fabrication methods, have been broadly reviewed. The last section of this review provides the introduction to alternative materials (Non-Zr) to Zr-based alloys for LWRs, such as advanced steels, Mo-based, and SiC-based materials.

Current status of materials development of nuclear fuel cladding tubes for light water reactors

Energy Technology Data Exchange (ETDEWEB)

Duan, Zhengang, E-mail: duan_zg@imr.tohoku.ac.jp [Department of Quantum Science and Energy Engineering, Graduate School of Engineering, Tohoku University, Sendai, Miyagi 980-8577 (Japan); Yang, Huilong [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan); Satoh, Yuhki [Institute for Materials Research, Tohoku University, Sendai, Miyagi 980-8577 (Japan); Murakami, Kenta; Kano, Sho; Zhao, Zishou; Shen, Jingjie [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan); Abe, Hiroaki, E-mail: abe.hiroaki@n.t.u-tokyo.ac.jp [Department of Nuclear Engineering, School of Engineering, The University of Tokyo, Nakagun, Ibaraki 319-1188 (Japan)

2017-05-15

Zirconium-based (Zr-based) alloys have been widely used as materials for the key components in light water reactors (LWRs), such as fuel claddings which suffer from waterside corrosion, hydrogen uptakes and strength loss at elevated temperature, especially during accident scenarios like the lost-of-coolant accident (LOCA). For the purpose of providing a safer, nuclear leakage resistant and economically viable LWRs, three general approaches have been proposed so far to develop the accident tolerant fuel (ATF) claddings: optimization of metallurgical composition and processing of Zr-based alloys, coatings on existing Zr-based alloys and replacement of current Zr-based alloys. In this manuscript, an attempt has been made to systematically present the historic development of Zr-based cladding, including the impacts of alloying elements on the material properties. Subsequently, the research investigations on coating layer on the surface of Zr-based claddings, mainly referring coating materials and fabrication methods, have been broadly reviewed. The last section of this review provides the introduction to alternative materials (Non-Zr) to Zr-based alloys for LWRs, such as advanced steels, Mo-based, and SiC-based materials.

Database structure and file layout of Nuclear Power Plant Database. Database for design information on Light Water Reactors in Japan

International Nuclear Information System (INIS)

Yamamoto, Nobuo; Izumi, Fumio.

1995-12-01

The Nuclear Power Plant Database (PPD) has been developed at the Japan Atomic Energy Research Institute (JAERI) to provide plant design information on domestic Light Water Reactors (LWRs) to be used for nuclear safety research and so forth. This database can run on the main frame computer in the JAERI Tokai Establishment. The PPD contains the information on the plant design concepts, the numbers, capacities, materials, structures and types of equipment and components, etc, based on the safety analysis reports of the domestic LWRs. This report describes the details of the PPD focusing on the database structure and layout of data files so that the users can utilize it efficiently. (author)

Status of advanced light water reactor designs 2004

International Nuclear Information System (INIS)

2004-05-01

The report is intended to be a source of reference information for interested organizations and individuals. Among them are decision makers of countries considering implementation of nuclear power programmes. Further, the report is addressed to government officials with an appropriate technical background and to research institutes of countries with existing nuclear programmes that wish to be informed on the global status in order to plan their nuclear power programmes including both research and development efforts and means for meeting future. The future utilization of nuclear power worldwide depends primarily on the ability of the nuclear community to further improve the economic competitiveness of nuclear power plants while meeting stringent safety requirements. The IAEA's activities in nuclear power technology development include the preparation of status reports on advanced reactor designs to provide all interested IAEA Member States with balanced and objective information on advances in nuclear plant technology. In the field of light water reactors, the last status report published by the IAEA was 'Status of Advanced Light Water Cooled Reactor Designs: 1996' (IAEA-TECDOC-968). Since its publication, quite a lot has happened: some designs have been taken into commercial operation, others have achieved significant steps toward becoming commercial products, including certification from regulatory authorities, some are in a design optimization phase to reduce capital costs, development for other designs began after 1996, and a few designs are no longer pursued by their promoters. With this general progress in mind, on the advice and with the support of the IAEA Department of Nuclear Energy's Technical Working Group on Advanced Technologies for Light Water Reactors (LWRs), the IAEA has prepared this new status report on advanced LWR designs that updates IAEA-TECDOC-968, presenting the various advanced LWR designs in a balanced way according to a common outline

Advanced reactor design study. Assessing nonbackfittable concepts for improving uranium utilization in light water reactors

International Nuclear Information System (INIS)

Fleischman, R.M.; Goldsmith, S.; Newman, D.F.; Trapp, T.J.; Spinrad, B.I.

1981-09-01

The objective of the Advanced Reactor Design Study (ARDS) is to identify and evaluate nonbackfittable concepts for improving uranium utilization in light water reactors (LWRs). The results of this study provide a basis for selecting and demonstrating specific nonbackfittable concepts that have good potential for implementation. Lead responsibility for managing the study was assigned to the Pacific Northwest Laboratory (PNL). Nonbackfittable concepts for improving uranium utilization in LWRs on the once-through fuel cycle were selected separately for PWRs and BWRs due to basic differences in the way specific concepts apply to those plants. Nonbackfittable concepts are those that are too costly to incorporate in existing plants, and thus, could only be economically incorporated in new reactor designs or plants in very early stages of construction. Essential results of the Advanced Reactor Design Study are summarized

Initial experimental evaluation of crud-resistant materials for light water reactors

Science.gov (United States)

Dumnernchanvanit, I.; Zhang, N. Q.; Robertson, S.; Delmore, A.; Carlson, M. B.; Hussey, D.; Short, M. P.

2018-01-01

The buildup of fouling deposits on nuclear fuel rods, known as crud, continues to challenge the worldwide fleet of light water reactors (LWRs). Crud causes serious operational problems for LWRs, including axial power shifts, accelerated fuel clad corrosion, increased primary circuit radiation dose rates, and in some instances has led directly to fuel failure. Numerous studies continue to attempt to model and predict the effects of crud, but each assumes that it will always be present. In this study, we report on the development of crud-resistant materials as fuel cladding coatings, to reduce or eliminate these problems altogether. Integrated loop testing experiments at flowing LWR conditions show significantly reduced crud adhesion and surface crud coverage, respectively, for TiC and ZrN coatings compared to ZrO2. The loop testing results roughly agree with the London dispersion component of van der Waals force predictions, suggesting that they contribute most significantly to the adhesion of crud to fuel cladding in out-of-pile conditions. These results motivate a new look at ways of reducing crud, thus avoiding many expensive LWR operational issues.

Characterization of LWRS Hybrid SiC-CMC-Zircaloy-4 Fuel Cladding after Gamma Irradiation

Energy Technology Data Exchange (ETDEWEB)

Isabella J van Rooyen

2012-09-01

The purpose of the gamma irradiation tests conducted at the Idaho National Laboratory (INL) was to obtain a better understanding of chemical interactions and potential changes in microstructural properties of a mock-up hybrid nuclear fuel cladding rodlet design (unfueled) in a simulated PWR water environment under irradiation conditions. The hybrid fuel rodlet design is being investigated under the Light Water Reactor Sustainability (LWRS) program for further development and testing of one of the possible advanced LWR nuclear fuel cladding designs. The gamma irradiation tests were performed in preparation for neutron irradiation tests planned for a silicon carbide (SiC) ceramic matrix composite (CMC) zircaloy-4 (Zr-4) hybrid fuel rodlet that may be tested in the INL Advanced Test Reactor (ATR) if the design is selected for further development and testing

Current status and future perspective of upgrading on light water reactors in Japan

International Nuclear Information System (INIS)

Tateki, Hoshion

1991-01-01

Energy demands in the world would be expected to increase constantly furthermore. In this situation the role of nuclear energy has become more important than it used to be from the view of the countermeasure for the global warming issue resulting from the exhausted gases causing the greenhouse effect in addition to the view of the energy securities. In also Japan, the energy demand-supply perspective plan issued on October,1990 which aims to expand the capacity of nuclear power plants to 72,500MWe as a target. by 2010. (see table.l) Present Light Water Reactors(LWRs) have been experienced and this type of reactor will remain as major stream of nuclear power generation for a long time all over the world. Considering above perspective, Advisory Committee for Energy of MITI(Ministry of International Trade and Industry) has investigated the technology and technical development of coming LWRs and concluded the report on June of this year. In this paper the contents of above report are introduced

Microencapsulated fuel technology for commercial light water and advanced reactor application

International Nuclear Information System (INIS)

Terrani, Kurt A.; Snead, Lance L.; Gehin, Jess C.

2012-01-01

The potential application of microencapsulated fuels to light water reactors (LWRs) has been explored. The specific fuel manifestation being put forward is for coated fuel particles embedded in silicon carbide or zirconium metal matrices. Detailed descriptions of these concepts are presented, along with a review of attributes, potential benefits, and issues with respect to their application in LWR environments, specifically from the standpoints of materials, neutronics, operations, and economics. Preliminary experiment and modeling results imply that with marginal redesign, significant gains in operational reliability and accident response margins could be potentially achieved by replacing conventional oxide-type LWR fuel with microencapsulated fuel forms.

A new book : 'light-water reactor materials'

International Nuclear Information System (INIS)

Olander, Donald R.; Motta, Arthur T.

2005-01-01

The contents of a new book currently in preparation are described. The dearth of books in the field of nuclear materials has left both students in nuclear materials classes and professionals in the same field without a resource for the broad fundamentals of this important sub-discipline of nuclear engineering. The new book is devoted entirely to materials problems in the core of light-water reactors, from the pressure vessel into the fuel. Key topics deal with the UO 2 fuel, zircaloy cladding, stainless steel, and of course, water. The restriction to LWR materials does not mean a short monograph; the enormous quantity of experimental and theoretical work over the past 50 years on these materials presents a challenge of culling the most important features and explaining them in the simplest quantitative fashion. Moreover, LWRs will probably be the sole instrument of the return of nuclear energy in electric power production for the next decade or so. By that time, a new book will be needed

Neutronic and Logistic Proposal for Transmutation of Plutonium from Spent Nuclear Fuel as Mixed-Oxide Fuel in Existing Light Water Reactors

International Nuclear Information System (INIS)

Trellue, Holly R.

2004-01-01

The use of light water reactors (LWRs) for the destruction of plutonium and other actinides [especially those in spent nuclear fuel (SNF)] is being examined worldwide. One possibility for transmutation of this material is the use of mixed-oxide (MOX) fuel, which is a combination of uranium and plutonium oxides. MOX fuel is used in nuclear reactors worldwide, so a large experience base for its use already exists. However, to limit implementation of SNF transmutation to only a fraction of the LWRs in the United States with a reasonable number of license extensions, full cores of MOX fuel probably are required. This paper addresses the logistics associated with using LWRs for this mission and the design issues required for full cores of MOX fuel. Given limited design modifications, this paper shows that neutronic safety conditions can be met for full cores of MOX fuel with up to 8.3 wt% of plutonium

Occupational radiation dose statistics from light-water power reactors operating in Western Europe

International Nuclear Information System (INIS)

Brookes, I.R.; Eng, T.

1987-01-01

Since the early days of nuclear power, collective and individual doses for people engaged in the maintenance and operation of nuclear power plants have been published by regulatory authorities. In 1979 a small working party whose members were drawn from Member States operating light-water reactors (LWRs) in the European Community was convened. The working party decided that only by collection of data under a unified scheme would it ever be possible to properly compare plant performance and for this reason a questionnaire was drawn up which attempted to elicit the maximum of information with the minimum inconvenience to the plant staff. Another decision made by the working party was to broaden the data base from 'European Community LWRs' to 'West European LWRs' to try to take advantage of the considerable experience being built up in Sweden, in Finland and in Switzerland. All the data available to the Commission up to the end of 1984 are presented and commented on. The deductions are not exhaustive but are believed to represent the limits of what could sensibly be done with the data available. Results are presented separately for BWR and PWR but no other subdivision, say by country or maker, is made. Where interpretation can be enhanced by graphical presentation, this is done. In general, doses for each job category are expressed in various ways to reveal and afford comparisons

Light Water Reactor Sustainability Program. Digital Architecture Requirements

Energy Technology Data Exchange (ETDEWEB)

Thomas, Kenneth [Idaho National Lab. (INL), Idaho Falls, ID (United States); Oxstrand, Johanna [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-03-01

The Digital Architecture effort is a part of the Department of Energy (DOE) sponsored Light-Water Reactor Sustainability (LWRS) Program conducted at Idaho National Laboratory (INL). The LWRS program is performed in close collaboration with industry research and development (R&D) programs that provides the technical foundations for licensing and managing the long-term, safe, and economical operation of current nuclear power plants (NPPs). One of the primary missions of the LWRS program is to help the U.S. nuclear industry adopt new technologies and engineering solutions that facilitate the continued safe operation of the plants and extension of the current operating licenses. Therefore, a major objective of the LWRS program is the development of a seamless digital environment for plant operations and support by integrating information from plant systems with plant processes for nuclear workers through an array of interconnected technologies. In order to get the most benefits of the advanced technology suggested by the different research activities in the LWRS program, the nuclear utilities need a digital architecture in place to support the technology. A digital architecture can be defined as a collection of information technology (IT) capabilities needed to support and integrate a wide-spectrum of real-time digital capabilities for nuclear power plant performance improvements. It is not hard to imagine that many processes within the plant can be largely improved from both a system and human performance perspective by utilizing a plant wide (or near plant wide) wireless network. For example, a plant wide wireless network allows for real time plant status information to easily be accessed in the control room, field workers’ computer-based procedures can be updated based on the real time plant status, and status on ongoing procedures can be incorporated into smart schedules in the outage command center to allow for more accurate planning of critical tasks. The goal

Once-through cycle, supercritical-pressure light water cooled reactor concept

Energy Technology Data Exchange (ETDEWEB)

Oka, Y.; Koshizuka, S. [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab

2001-07-01

Concept of once-through cycle, supercritical-pressure light water cooled reactors was developed. The research covered major aspects of conceptual design such as cores of thermal and fast reactors, plant system and heat balance, safety system and criteria, accident and transient analysis, LOCA, PSA, plant control and start-up. The advantages of the reactor lie in the compactness of the plant from high specific enthalpy of supercritical water, the simplicity of the once-through cycle and the experiences of major component technologies which are based on supercritical fossil-fired power plants and LWRs. The operating temperatures of the major components are within the experience in spite of high coolant outlet temperature. The once-through cycle is compatible with the tight fuel lattice fast reactor because of high head pumps and small coolant flow rate. (author)

Once-through cycle, supercritical-pressure light water cooled reactor concept

International Nuclear Information System (INIS)

Oka, Y.; Koshizuka, S.

2001-01-01

Concept of once-through cycle, supercritical-pressure light water cooled reactors was developed. The research covered major aspects of conceptual design such as cores of thermal and fast reactors, plant system and heat balance, safety system and criteria, accident and transient analysis, LOCA, PSA, plant control and start-up. The advantages of the reactor lie in the compactness of the plant from high specific enthalpy of supercritical water, the simplicity of the once-through cycle and the experiences of major component technologies which are based on supercritical fossil-fired power plants and LWRs. The operating temperatures of the major components are within the experience in spite of high coolant outlet temperature. The once-through cycle is compatible with the tight fuel lattice fast reactor because of high head pumps and small coolant flow rate. (author)

A liquid-metal reactor for burning minor actinides of spent light water reactor fuel. 1: Neutronics design study

International Nuclear Information System (INIS)

Choi, H.; Downar, T.J.

1999-01-01

A liquid-metal reactor was designed for the primary purpose of burning the minor actinide waste from commercial light water reactors (LWRs). The design was constrained to maintain acceptable safety performance as measured by the burnup reactivity swing, the Doppler constant, and the sodium void worth. Sensitivity studies were performed for homogeneous and decoupled core designs, and a minor actinide burner design was determined to maximize actinide consumption and satisfy safety constraints. One of the principal innovations was the use of two core regions, with a fissile plutonium outer core and an inner core consisting only of minor actinides. The physics studies performed here indicate that a 1200-MW(thermal) core is able to consume the annual minor actinide inventory of about 16 LWRs and still exhibit reasonable safety characteristics

Nondestructive examination (NDE) reliability for inservice inspection of light water reactors

International Nuclear Information System (INIS)

Doctor, S.R.; Deffenbaugh, J.D.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.

1989-10-01

The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvement that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements based on material properties, service conditions, and NDE capabilities and uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other inspected components. This is a progress report covering the programmatic work from October 1987 through March 1988. 21 refs., 28 figs., 2 tabs

Nondestructive examination (NDE) reliability for inservice inspection of light waters reactors

International Nuclear Information System (INIS)

Doctor, S.R.; Deffenbaugh, J.D.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.

1989-11-01

Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other inspected components. This is a progress report covering the programmatic work from April 1988 through September 1988. 33 refs., 70 figs., 12 tabs

Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors

International Nuclear Information System (INIS)

Doctor, S.R.; Diaz, A.A.; Friley, J.R.

1993-09-01

The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other components inspected in accordance with Section XI of the ASME Code. This is a progress report covering the programmatic work from October 1991 through March 1992

Transmutation of Americium in Light and Heavy Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Hyland, B.; Dyck, G.R.; Edwards, G.W.R. [Chalk River Laboratories, Atomic Energy of Canada Limited (Canada); Ellis, R.J.; Gehin, J.C. [Oak Ridge National Laboratory (ORNL), Oak Ridge, Tennessee (United States); Maldonado, G.I. [University of Tennessee (Knoxville)/ORNL, Tennessee (United States)

2009-06-15

There is interest worldwide in reducing the burden on geological nuclear fuel disposal sites. In most disposal scenarios the decay heat loading of the surrounding rock limits the capacity of these sites. On the long term, this decay heat is generated primarily by actinides, and a major contributor 100 to 1000 years after discharge from the reactor is {sup 241}Am. One possible approach to reducing the decay-heat burden is to reprocess spent reactor fuel and use thermal spectrum reactors to 'burn' the Am nuclides. The viability of this approach is dependent upon the detailed changes in chemical and isotopic composition of actinide-bearing fuels after irradiation in thermal reactor spectra. The currently available thermal spectrum reactor options include light water-reactors (LWRs) and heavy-water reactors (HWRs) such as the CANDU{sup R} designs. In addition, as a result of the recycle of spent LWR fuel, there would be a considerable amount of potential recycled uranium (RU). One proposed solution for the recycled uranium is to use it as fuel in Candu reactors. This paper investigates the possibilities of transmuting americium in 'spiked' bundles in pressurized water reactors (PWRs) and in boiling water reactors (BWRs). Transmutation of Am in Candu reactors is also examined. One scenario studies a full core fuelled with homogeneous bundles of Am mixed with recycled uranium, while a second scenario places Am in an inert matrix in target channels in a Candu reactor, with the rest of the reactor fuelled with RU. A comparison of the transmutation in LWRs and HWRs is made, in terms of the fraction of Am that is transmuted and the impact on the decay heat of the spent nuclear fuel. CANDU{sup R} is a registered trademark of Atomic Energy of Canada Limited (AECL). (authors)

Environmentally assisted cracking in Light Water Reactors: Semiannual report, April 1993--September 1993. Volume 17

International Nuclear Information System (INIS)

Chopra, O.K.; Chung, H.M.; Karlsen, T.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K.

1994-06-01

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) during the six months from April 1993 to September 1993. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels; (b) EAC of cast stainless steels (SSs); and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking of Type 304 SS after accumulation of relatively high fluence. Fatigue tests were conducted on medium-sulfur-content A106-Gr B piping and A533-Gr B pressure vessel steels in simulated PWR water and in air. Additional crack growth data were obtained on fracture-mechanics specimens of cast austenitic SSs in the as-received and thermally aged conditions in simulated boiling-water reactor (BWR) water at 289 degree C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section 11 of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating BWRs were studied by Auger electron spectroscopy and scanning electron microscopy

Fuel -coolant interactions in LWRs and LMFBRs: relationships and distinctions

Energy Technology Data Exchange (ETDEWEB)

Duffey, R B; Lellouche, G S [Nuclear Safety and Analysis Department, Electric Power Research Institute, Palo Alto, CA (United States)

1979-10-15

The question of fuel-coolant interaction and of potential vapor explosion is raised here. lt is the contention of the authors that there is in fact no need to study this question vis a vis Light Water Reactors (LWR) except from an academic point of view since it does not impact on safety considerations. As for LMFBRs, the design basis whole core accidents for LWRs are derived from the fundamental concern of maintaining core geometry to provide for convective cooling. However, the important distinction is that the core is in its most reactive configuration, and core and fuel rearrangement is therefore not of such concern. The author's thesis is that even if the probability of steam explosion following core melt were two orders of magnitude greater than currently assumed (10{sup -2}) the total LWR risk would increase only by a factor of 2-6 for BWRs and less a factor of 10 for PWRs

Guidance for Developing Principal Design Criteria for Advanced (Non-Light Water) Reactors

Energy Technology Data Exchange (ETDEWEB)

Holbrook, Mark [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, Jim [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2015-03-01

In July 2013, the US Department of Energy (DOE) and US Nuclear Regulatory Commission (NRC) established a joint initiative to address a key portion of the licensing framework essential to advanced (non-light water) reactor technologies. The initiative addressed the “General Design Criteria for Nuclear Power Plants,” Appendix A to10 Code of Federal Regulations (CFR) 50, which were developed primarily for light water reactors (LWRs), specific to the needs of advanced reactor design and licensing. The need for General Design Criteria (GDC) clarifications in non-LWR applications has been consistently identified as a concern by the industry and varied stakeholders and was acknowledged by the NRC staff in their 2012 Report to Congress1 as an area for enhancement. The initiative to adapt GDC requirements for non-light water advanced reactor applications is being accomplished in two phases. Phase 1, managed by DOE, consisted of reviews, analyses and evaluations resulting in recommendations and deliverables to NRC as input for NRC staff development of regulatory guidance. Idaho National Laboratory (INL) developed this technical report using technical and reactor technology stakeholder inputs coupled with analysis and evaluations provided by a team of knowledgeable DOE national laboratory personnel with input from individual industry licensing consultants. The DOE national laboratory team reviewed six different classes of emerging commercial reactor technologies against 10 CFR 50 Appendix A GDC requirements and proposed guidance for their adapted use in non-LWR applications. The results of the Phase 1 analysis are contained in this report. A set of draft Advanced Reactor Design Criteria (ARDC) has been proposed for consideration by the NRC in the establishment of guidance for use by non-LWR designers and NRC staff. The proposed criteria were developed to preserve the underlying safety bases expressed by the original GDC, and recognizing that advanced reactors may take

Status and subjects of thermal-hydraulic analysis for next-generation LWRs with passive safety systems

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-03-01

The present status and subjects on thermal-hydraulic analysis for next-generation light water reactors (LWRs) with passive safety systems were summarized based on survey results and discussion by subcommittee on improvement of reactor thermal-hydraulic analysis codes under nuclear code committee in Japan Atomic Energy Research Institute. This survey was performed to promote the research of improvement of reactor thermal-hydraulic analysis codes in future. In the first part of this report, the status and subjects on system analysis and those on evaluation of passive safety system performance are summarized for various types of reactor proposed before. In the second part, the status and subjects on multidimensional two-phase flow analysis are reviewed, since the multidimensional analysis was recognized as one of most important subjects through the investigation in the first part. Besides, databases for bubbly flow and annular dispersed flow were explored, those are needed to assess and verify each multidimensional analytical method. The contents in this report are the forefront of thermal-hydraulic analysis for LWRs and those include current findings for the development of multidimensional two-phase flow analytical method. Thus, we expect that the contents can offer various useful information against the improvement of reactor thermal-hydraulic analysis codes in future. (author)

Flow-induced vibration for light water reactors. Progress report, January-June 1980

International Nuclear Information System (INIS)

De Coster, M.A.

1981-02-01

Flow-Induced Vibration for Light Water Reactors (FIV for LWRs) is a four-year program designed to improve the FIV performance of light water reactors through the development of design criteria, analytical models for predicting behavior of components, general scaling laws to improve the accuracy of reduced-scale tests, and the identification of high FIV risk areas. The program is managed by the General Electric Nuclear Power Systems Engineering Department and has three major contributors: General Electric Nuclear Power Systems Engineering Department (NPSED), General Electric Corporate Research and Development (CR and D) and Argonne National Laboratory (ANL). The program commenced December 1, 1976, but was suspended on September 30, 1978, due to a shift in Department of Energy (DOE) priorities away from LWR productivity/availability. It was reinitiated as of August 1, 1979. A second program suspension occurred from March 29, 1980 through May 16, 1980, due to funding limits. This progress report summarizes the accomplishments achieved during the period from Janury 1980 to June 1980

Performance of Transuranic-Loaded Fully Ceramic Micro-Encapsulated Fuel in LWRs Final Report, Including Void Reactivity Evaluation

International Nuclear Information System (INIS)

Pope, Michael A.; Sen, R. Sonat; Boer, Brian; Ougouag, Abderrafi M.; Youinou, Gilles

2011-01-01

The current focus of the Deep Burn Project is on once-through burning of transuranics (TRU) in light-water reactors (LWRs). The fuel form is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the tri-isotropic (TRISO) fuel particle design from high-temperature reactor technology. In the Deep Burn LWR (DB-LWR) concept, these fuel particles are pressed into compacts using SiC matrix material and loaded into fuel pins for use in conventional LWRs. The TRU loading comes from the spent fuel of a conventional LWR after 5 years of cooling. Unit cell and assembly calculations have been performed using the DRAGON-4 code to assess the physics attributes of TRU-only FCM fuel in an LWR lattice. Depletion calculations assuming an infinite lattice condition were performed with calculations of various reactivity coefficients performed at each step. Unit cells and assemblies containing typical UO2 and mixed oxide (MOX) fuel were analyzed in the same way to provide a baseline against which to compare the TRU-only FCM fuel. Then, assembly calculations were performed evaluating the performance of heterogeneous arrangements of TRU-only FCM fuel pins along with UO2 pins.

Environmentally assisted cracking in light water reactors

International Nuclear Information System (INIS)

Chopra, O.K.; Chung, H.M.; Gruber, E.E.

1996-07-01

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from April 1995 to December 1995. Topics that have been investigated include fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, EAC of Alloy 600 and 690, and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water that contained various concentrations of dissolved oxygen (DO) to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Crack-growth-rate tests were conducted on compact-tension specimens from several heats of Alloys 600 and 690 in simulated LWR environments. Effects of fluoride-ion contamination on susceptibility to intergranular cracking of high- and commercial- purity Type 304 SS specimens from control-tensile tests at 288 degrees Centigrade. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials

Evaluation of Proactive Management Issues Associated with Materials Aging in Light Water Reactors

International Nuclear Information System (INIS)

Shoji, T.; Takeda, Y.; Kuniya, J.

2012-01-01

A Proactive Materials Degradation Management (PMDM) project has been carried out at the Fracture Research Institute (FRI), Tohoku University for 4 years, as a part of a Nuclear Industries Safety Agency (NISA) project that was formed in 2007 to define an Aging Management Program that addresses unexpected structural material failures in Light Water Reactors (LWRs). Such a program required, therefore, the development of a life prediction capability for specific combinations of degradation modes, structural materials, and reactor components. In this paper, the research subjects needed to predict quantitatively aging degradation phenomena in LWR structural materials are introduced. (author)

Controlling radiation fields in siemans designed light water reactors

Energy Technology Data Exchange (ETDEWEB)

Riess, R.; Marchl, T. [Siemens Power Generation Group, Erlangen (Germany)

1995-03-01

An essential item for the control of radiation fields is the minimization of the use of satellites in the reactor systems of Light Water Reactors (LWRs). A short description of the qualification of Co-replacement materials will be followed by an illustration of the locations where these materials were implemented in Siemens designed LWRs. Especially experiences in PWRs show the immense influence of reduction of cobalt sources on dose rate buildup. The corrosion and the fatique and wear behavior of the replacement materials has not created concern up to now. A second tool to keep occupational radiation doses at a low level in PWRs is the use of the modified B/Li-chemistry. This is practized in Siemens designed plants by keeping the Li level at a max. value of 2 ppm until it reaches a pH (at 300{degrees}C) of {approximately}7.4. This pH is kept constant until the end of the cycle. The substitution of cobalt base alloys and thus the removal of the Co-59 sources from the system had the largest impact on the radiation levels. Nonetheless, the effectiveness of the coolant chemistry should not be neglected either. Several years of successful operation of PWRs with the replacement materials resulted in an occupational radiation exposure which is below 0.5 man-Sievert/plant and year.

Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors

International Nuclear Information System (INIS)

Doctor, S.R.; Diaz, A.A.; Friley, J.R.; Greenwood, M.S.; Heasler, P.G.; Kurtz, R.J.; Simonen, F.A.; Spanner, J.C.; Vo, T.V.

1993-11-01

The Evaluation and Improvement of NDE Reliability for Inservice inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs);using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to the Regulatory and ASME Code requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel and other components inspected in accordance with Section XI of the ASME Code. This is a programs report covering the programmatic work from April 1992 through September 1992

Nondestructive examination (NDE) reliability for inservice inspection of light water reactors

International Nuclear Information System (INIS)

Doctor, S.R.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Vo, T.V.

1991-08-01

The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other components inspected in accordance with Section 6 of the ASME Code. This is a progress report covering the pro grammatic work from April 1989 through September 1989. 12 refs., 4 figs. 5 tabs

Hydrogen considerations in light-water power reactons

International Nuclear Information System (INIS)

Keilholtz, G.W.

1976-02-01

A critical review of the literature now available on hydrogen considerations in light-water power reactors (LWRs) and a bibliography of that literature are presented. The subject matter includes mechanisms for the generation of hydrogen-oxygen mixtures, a description of the fundamental properties of such mixtures, and their spontaneous ignition in both static and dynamic systems. The limits for hydrogen flammability and flame propagation are examined in terms of the effects of pressure, temperature, and additives; the emphasis is on the effects of steam and water vapor. The containment systems for pressurized-water reactors (PWRs) and boiling-water reactors (BWRs) are compared, and methods to control hydrogen and oxygen under the conditions of both normal operation and postulated accidents are reviewed. It is concluded that hydrogen can be controlled so that serious complications from the production of hydrogen will not occur. The bibliography contains abstracts from the computerized files of the Nuclear Safety Information Center. Key-word, author, and permuted-title indexes are provided. The bibliography includes responses to questions asked by the U. S. Nuclear Regulatory Commission (NRC) which relate to hydrogen, as well as information on normal operations and postulated accidents including generation of hydrogen from core sprays. Other topics included in the ten sections of the bibliography are metal-water reactions, containment atmosphere, radiolytic gas, and recombiners

Closed ThUOX Fuel Cycle for LWRs with ADTT (ATW) Backend for the 21st Century

International Nuclear Information System (INIS)

Beller, D.E.; Sailor, W.C.; Venneri, F.

1998-01-01

A future nuclear energy scenario with a closed, thorium-uranium-oxide (ThUOX) fuel cycle and new light water reactors (TULWRs) supported by Accelerator Transmutation of Waste (ATW) systems could provide several improvements beyond today's once-through, UO 2 -fueled nuclear technology. A deployment scenario with TULWRs plus ATWs to burn the actinides produced by these LWRs and to close the back-end of the ThUOX fuel cycle was modeled to satisfy a US demand that increases linearly from 80 GWe in 2020 to 200 GWe by 2100. During the first 20 years of the scenario (2000-2020), nuclear energy production in the US declines from today's 100 GWe to about 80 GWe, in accordance with forecasts of the US DOE's Energy Information Administration. No new nuclear systems are added during this declining nuclear energy period, and all existing LWRs are shut down by 2045. Beginning in 2020, ATWs that transmute the actinides from existing LWRs are deployed, along with TULWRs and additional ATWs with a support ratio of 1 ATW to 7 TULWRs to meet the energy demand scenario. A final mix of 174 GWe from TULWRs and 26 GWe from ATWs provides the 200 GWe demand in 2100. Compared to a once-through LWR scenario that meets the same energy demand, the TULWR/ATW concept could result in the following improvements: depletion of natural uranium resources would be reduced by 50%; inventories of Pu which may result in weapons proliferation will be reduced in quantity by more than 98% and in quality because of higher neutron emissions and 50 times the alpha-decay heating of weapons-grade plutonium; actinides (and possibly fission products) for final disposal in nuclear waste would be substantially reduced; and the cost of fuel and the fuel cycle may be 20-30% less than the once-through UO 2 fuel cycle

Perspectives on phenomenology and simulation of severe accident in light water reactors

International Nuclear Information System (INIS)

Sugimoto, Jun

2014-01-01

Severe accident phenomena in light water reactors (LWRs) are generally characterized by their physically and chemically complex processes involved with high temperature core melt, multi-component and multi-phase flows, transport of radioactive materials and sometimes highly non-equilibrium state. Severe accident phenomenology is usually categorized into four phases; (1) fuel degradation, (2) in-vessel phenomena, (3) ex-vessel phenomena and (4) fission product release and transport. Among these, ex-vessel phenomena consist of five subcategories; 1) direct containment heating, 2) fuel coolant interaction (steam explosion), 3) molten core concrete interaction, 4) hydrogen behaviour and control and 5) containment failure/leakage. In the field of simulation of severe accident, severe accident analytical codes have been developed in the United States, EU and Japan, such as MAAP, MELCOR, ASTEC, THALES and SAMPSON. Many different kinds of analytical codes for the specific severe accident phenomena have also been developed worldwide. After the accident at Fukushima Daiichi Nuclear Power Station, review of severe accident research issues has been conducted and several issues are reconsidered, such as effects of BWR core degradation behaviors, sea water injection, pool scrubbing under rapid depressurization, containment failure/leakage and re-criticality. Some new experimental and analytical efforts have been started after the Fukushima accident. The present paper describes the perspectives on phenomenology and simulation of severe accident in LWRs, with the emphasis of insights obtained in the review of Fukushima accident. (author)

A functional method for estimating DPA tallies in Monte Carlo calculations of Light Water Reactors

International Nuclear Information System (INIS)

Read, Edward A.; Oliveira, Cassiano R.E. de

2011-01-01

There has been a growing need in recent years for the development of methodology to calculate radiation damage factors, namely displacements per atom (dpa), of structural components for Light Water Reactors (LWRs). The aim of this paper is to discuss the development and implementation of a dpa method using Monte Carlo method for transport calculations. The capabilities of the Monte Carlo code Serpent such as Woodcock tracking and fuel depletion are assessed for radiation damage calculations and its capability demonstrated and compared to those of the Monte Carlo code MCNP for radiation damage calculations of a typical LWR configuration. (author)

Oxidation of SiC cladding under Loss of Coolant Accident (LOCA) conditions in LWRs

International Nuclear Information System (INIS)

Lee, Y.; Yue, C.; Arnold, R. P.; McKrell, T. J.; Kazimi, M. S.

2012-01-01

An experimental assessment of Silicon Carbide (SiC) cladding oxidation rate in steam under conditions representative of Loss of Coolant Accidents (LOCA) in light water reactors (LWRs) was conducted. SiC oxidation tests were performed with monolithic alpha phase tubular samples in a vertical quartz tube at a steam temperature of 1140 deg. C and steam velocity range of 1 to 10 m/sec, at atmospheric pressure. Linear weight loss of SiC samples due to boundary layer controlled reaction of silica scale (SiO 2 volatilization) was experimentally observed. The weight loss rate increased with increasing steam flow rate. Over the range of test conditions, SiC oxidation rates were shown to be about 3 orders of magnitude lower than the oxidation rates of zircaloy 4. A SiC volatilization correlation for developing laminar flow in a vertical channel is formulated. (authors)

Environmentally assisted cracking in light water reactors. Semiannual report, October 1993--March 1994. Volume 18

International Nuclear Information System (INIS)

Chung, H.M.; Chopra, O.K.; Erck, R.A.; Kassner, T.F.; Michaud, W.F.; Ruther, W.E.; Sanecki, J.E.; Shack, W.J.; Soppet, W.K.

1995-03-01

This report summarizes work performed by Argonne National Laboratory (ANL) on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) during the six months from October 1993 to March 1994. EAC and fatigue of piping, pressure vessels, and core components in LWRs are important concerns in operating plants and as extended reactor lifetimes are envisaged. Topics that have been investigated include (a) fatigue of low-alloy steel used in piping, steam generators, and reactor pressure vessels, (b) EAC of wrought and cast austenitic stainless steels (SSs), and (c) radiation-induced segregation and irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS after accumulation of relatively high fluence. Fatigue tests have been conducted on A302-Gr B low-alloy steel to verify whether the current predictions of modest decreases of fatigue life in simulated pressurized water reactor water are valid for high-sulfur heats that show environmentally enhanced fatigue crack growth rates. Additional crack growth data were obtained on fracture-mechanics specimens of austenitic SSs to investigate threshold stress intensity factors for EAC in high-purity oxygenated water at 289 degrees C. The data were compared with predictions based on crack growth correlations for wrought austenitic SS in oxygenated water developed at ANL and rates in air from Section XI of the ASME Code. Microchemical and microstructural changes in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath from operating boiling water reactors were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements, which are not specified in the ASTM specifications, may contribute to IASCC of solution-annealed materials

Evaluation and improvement in nondestructive examination (NDE) reliability for inservice inspection of light water reactors

International Nuclear Information System (INIS)

Doctor, S.R.; Deffenbaugh, J.D.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.

1988-01-01

The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) program at the Pacific Northwest Laboratory was established by the NRC to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other inspected components. This is a progress report covering the programmatic work from October 1986 through September 1987. (author)

Evaluation and improvement in nondestructive examination (NDE) reliability for inservice inspection of light water reactors

International Nuclear Information System (INIS)

Doctor, S.R.; Deffenbaugh, J.D.; Good, M.S.; Green, E.R.; Heasler, P.G.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.

1988-01-01

The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactor (NDE Reliability) program at the Pacific Northwest Laboratory was established by the NRC to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWRs); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to ASME Code and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other inspected components. This is a progress report covering the programmatic work from October 1986 through September 1987

Solar energy as an alternate energy source to mixed oxide fuels in light-water cooled reactors

International Nuclear Information System (INIS)

Bertini, H.W.

1977-01-01

Supplemental information pertaining to the generic environmental impact statement on the Pu recycling process for mixed oxide light-water cooled reactors (GESMO) was requested from several sources. In particular, the role of alternate sources of energy was to be explored and the implications of these alternate sources to the question of Pu recycle in LWRs were to be investigated. In this vein, solar energy as an alternate source is the main subject of this report, along with other information related to solar energy. The general conclusion is that solar energy should have little effect on the decisions concerning GESMO

On Monte Carlo estimation of radiation damage in light water reactor systems

International Nuclear Information System (INIS)

Read, Edward A.; Oliveira, Cassiano R.E. de

2010-01-01

There has been a growing need in recent years for the development of methodologies to calculate damage factors, namely displacements per atom (dpa), of structural components for Light Water Reactors (LWRs). The aim of this paper is discuss and highlight the main issues associated with the calculation of radiation damage factors utilizing the Monte Carlo method. Among these issues are: particle tracking and tallying in complex geometries, dpa calculation methodology, coupled fuel depletion and uncertainty propagation. The capabilities of the Monte Carlo code Serpent such as Woodcock tracking and burnup are assessed for radiation damage calculations and its capability demonstrated and compared to those of the MCNP code for dpa calculations of a typical LWR configuration involving the core vessel and the downcomer. (author)

Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

Energy Technology Data Exchange (ETDEWEB)

Cowell, B.S.; Fisher, S.E.

1999-02-01

The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option.

Survey of Worldwide Light Water Reactor Experience with Mixed Uranium-Plutonium Oxide Fuel

International Nuclear Information System (INIS)

Cowell, B.S.; Fisher, S.E.

1999-01-01

The US and the Former Soviet Union (FSU) have recently declared quantities of weapons materials, including weapons-grade (WG) plutonium, excess to strategic requirements. One of the leading candidates for the disposition of excess WG plutonium is irradiation in light water reactors (LWRs) as mixed uranium-plutonium oxide (MOX) fuel. A description of the MOX fuel fabrication techniques in worldwide use is presented. A comprehensive examination of the domestic MOX experience in US reactors obtained during the 1960s, 1970s, and early 1980s is also presented. This experience is described by manufacturer and is also categorized by the reactor facility that irradiated the MOX fuel. A limited summary of the international experience with MOX fuels is also presented. A review of MOX fuel and its performance is conducted in view of the special considerations associated with the disposition of WG plutonium. Based on the available information, it appears that adoption of foreign commercial MOX technology from one of the successful MOX fuel vendors will minimize the technical risks to the overall mission. The conclusion is made that the existing MOX fuel experience base suggests that disposition of excess weapons plutonium through irradiation in LWRs is a technically attractive option

Light water reactor safety. Past, present and future

International Nuclear Information System (INIS)

Sehgal, Bal Raj

2009-01-01

This paper presents a review of the past, present and possible future developments in light water reactor (LWR) safety. The paper divides the past into two periods: the distant past i.e., before the TMI-2 accident when the main concern was with the design basis, the general design criteria, the concept of the defense in depth, the thermal hydraulics of the large loss of coolant accident (LOCA) and the success of the emergency core cooling system (ECCS), and the near past, i.e., after the TMI-2 accident when the main concern was with the physics of the postulated severe accidents: their prevention and mitigation. The present period is chosen as the translation of the research on the design basis and severe accidents into practical designs of Gen III+ with their core catchers and severe accident management (SAM) strategies, which could, in fact, provide ample assurances of public safety even for very severe accidents. The paper attempts to describe the remaining safety issues for both the Gen II and Gen III+ nuclear plants. The more important safety challenges are being posed by the recent moves of (1) extension of the life of the presently installed Gen II LWRs to 60 years (and perhaps to 80 years) and (2) the large uprates in power that are being sought for the Gen II LWRs. Clearly, the safety margins will be tested by these moves of long extended operations with greater power ratings of the Gen II plants. A prognosis of the emerging development trends in the LWR safety has been attempted with some suggestions. (author)

Disposition of weapons-grade plutonium in LWRs - a utility perspective

International Nuclear Information System (INIS)

Naughton, W.F.

1996-01-01

The optimal alternative for the disposition of weapons-grade plutonium (WPu) is to burn it in currently operating light water reactors (LWRs). There are three key aspects associated with the reactor burn option that lead to this conclusion. They are timeliness, acceptable solution, and equity or resource recovery. The National Academy of Sciences report on this subject has thoroughly covered the first two aspects by indicating that there is open-quotes a clear and present dangerclose quotes associated with WPu - the timeliness issue. Also, the report indicates that the spent-fuel standard is the acceptable long-term solution. Both of these aspects are met by the reactor burn option as it exists today; i.e., the reactor option is both timely and obviously meets the spent-fuel standard. The equity or resource recovery aspect is based on the fact that the resources for the development and use of this material as a nuclear deterrent for the past 50 yr was supplied by, in the U.S.'s case, the taxpayers. Only the reactor burn option offers an opportunity for the partial recovery of those resources in the form of electrical energy

Fatigue and environmentally assisted cracking in light water reactors

International Nuclear Information System (INIS)

Kassner, T.F.; Ruther, W.E.; Chung, H.M.; Hicks, P.D.; Hins, A.G.; Park, J.Y.; Shack, W.J.

1991-12-01

Fatigue and environmentally assisted cracking of piping, pressure vessels, and core components in light water reactors (LWRs) are important concerns as extended reactor lifetimes are envisaged. The degradation processes include intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel (SS) piping in boiling water reactors (BWRs), and propagation of fatigue or SCC cracks (which initiate in sensitized SS cladding) into low-alloy ferritic steels in BWR pressure vessels. Similar cracking has also occurred in upper shell-to-transition cone girth welds in pressurized water reactor (PWR) steam generator vessels. Another concern is failure of reactor-core internal components after accumulation of relatively high fluence, which has occurred in both BWRs and PWRs. Research during the past year focused on (1) fatigue and SCC of ferritic steels used in piping and in steam generator and reactor pressure vessels, (2) role of chromate and sulfate in simulated BWR water in SCC of sensitized Type 304 SS, and (3) irradiation-assisted SCC in high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes used in two operating BWRs. Failure after accumulation of relatively high fluence has been attributed to radiation-induced segregation (RIS) of elements such as Si, P, Ni, and Cr. This document provides a summary of research progress in these areas

Oak Ridge National Laboratory Support of Non-light Water Reactor Technologies: Capabilities Assessment for NRC Near-term Implementation Action Plans for Non-light Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Belles, Randy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Jain, Prashant K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-04-01

The Oak Ridge National Laboratory (ORNL) has a rich history of support for light water reactor (LWR) and non-LWR technologies. The ORNL history involves operation of 13 reactors at ORNL including the graphite reactor dating back to World War II, two aqueous homogeneous reactors, two molten salt reactors (MSRs), a fast-burst health physics reactor, and seven LWRs. Operation of the High Flux Isotope Reactor (HFIR) has been ongoing since 1965. Expertise exists amongst the ORNL staff to provide non-LWR training; support evaluation of non-LWR licensing and safety issues; perform modeling and simulation using advanced computational tools; run laboratory experiments using equipment such as the liquid salt component test facility; and perform in-depth fuel performance and thermal-hydraulic technology reviews using a vast suite of computer codes and tools. Summaries of this expertise are included in this paper.

Comparison of rate theory based modeling calculations with the surveillance test results of Korean light water reactors

International Nuclear Information System (INIS)

Lee, Gyeong Geun; Lee, Yong Bok; Kim, Min Chul; Kwon, Junh Yun

2012-01-01

Neutron irradiation to reactor pressure vessel (RPV) steels causes a decrease in fracture toughness and an increase in yield strength while in service. It is generally accepted that the growth of point defect cluster (PDC) and copper rich precipitate (CRP) affects radiation hardening of RPV steels. A number of models have been proposed to account for the embrittlement of RPV steels. The rate theory based modeling mathematically described the evolution of radiation induced microstructures of ferritic steels under neutron irradiation. In this work, we compared the rate theory based modeling calculation with the surveillance test results of Korean Light Water Reactors (LWRs)

Evaluation of alternative partitioning/transmutation scenarios using transmutation in light-water reactors (LWRs)

International Nuclear Information System (INIS)

Collins, E.D.; Renier, J.P.; Del Cul, B.; Spencer, B.

2005-01-01

Previous Advanced Fuel Cycle Initiative (AFCI) studies were made to assess the effects of the existing accumulation of LWR spent fuel in the United States on the capability to partition/transmute actinides using existing and advanced LWR. The concept of treating the oldest fuel first indicated that significant advantages could be gained in both partitioning, transmutation, and in overall cost reduction. The processing scenarios previously evaluated assumed that (1) 2000 MT/year of spent fuel, irradiated to 45 GWd/MT and decayed for 30 years is processed; (2) recovered plutonium and 90% of the neptunium are transmuted in LWR MOX fuel; and (3) minor actinides, consisting of americium, curium, and 10% of the neptunium are transmuted in burnable poison type targets. Results of the previous study showed that significant benefits could be obtained, including (1) lower costs for partitioning and transmutation and for storage of spent fuel, (2) maintenance of proliferation resistance for the fissile plutonium in spent fuels, and (3) extended lifetime for the repository. The lower costs would be achieved primarily because no capital investment for a special transmuter reactor (fast reactor, accelerator-driven system, etc.) would be required. Instead, only existing and new LWRs would be utilized. Moreover, no new storage capacity could be needed for spent fuels and irradiated targets because the number of spent fuel assemblies would remain the same after the scenario is begun. Even though the total inventory of plutonium would rise during the early cycles, ∼98% of the plutonium would be contained in stored spent fuel and would be protected by high radiation (the Spent Fuel Standard). This is because the spent fuel would be reprocessed and re-irradiated at intervals within which the fission products, 137 Cs and 90 Sr, both with half-lives of ∼ 30 years, exist in significantly high concentrations.The lifetime of the repository would be extended significantly because all of

Light Water Reactor Sustainability Program Integrated Program Plan

International Nuclear Information System (INIS)

Griffith, George; Youngblood, Robert; Busby, Jeremy; Hallbert, Bruce; Barnard, Cathy; McCarthy, Kathryn

2012-01-01

Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline - even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy's Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration's energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans.

Light Water Reactor Sustainability Program Integrated Program Plan

Energy Technology Data Exchange (ETDEWEB)

George Griffith; Robert Youngblood; Jeremy Busby; Bruce Hallbert; Cathy Barnard; Kathryn McCarthy

2012-01-01

Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline - even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy's Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration's energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans.

Light Water Reactor Sustainability Program Integrated Program Plan

Energy Technology Data Exchange (ETDEWEB)

McCarthy, Kathryn A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Busby, Jeremy [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hallbert, Bruce [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bragg-Sitton, Shannon [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Barnard, Cathy [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-04-01

Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

Light Water Reactor Sustainability Program Integrated Program Plan

Energy Technology Data Exchange (ETDEWEB)

Kathryn McCarthy; Jeremy Busby; Bruce Hallbert; Shannon Bragg-Sitton; Curtis Smith; Cathy Barnard

2013-04-01

Nuclear power has safely, reliably, and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. Domestic demand for electrical energy is expected to experience a 31% growth from 2009 to 2035. At the same time, most of the currently operating nuclear power plants will begin reaching the end of their initial 20-year extension to their original 40-year operating license for a total of 60 years of operation. Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary in 2009. The U.S. Department of Energy Office of Nuclear Energy’s Research and Development Roadmap (Nuclear Energy Roadmap) organizes its activities around four objectives that ensure nuclear energy remains a compelling and viable energy option for the United States. The four objectives are as follows: (1) develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of the current reactors; (2) develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration’s energy security and climate change goals; (3) develop sustainable nuclear fuel cycles; and (4) understand and minimize the risks of nuclear proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans.

Light Water Reactor Sustainability Program A Reference Plan for Control Room Modernization: Planning and Analysis Phase

Energy Technology Data Exchange (ETDEWEB)

Jacques Hugo; Ronald Boring; Lew Hanes; Kenneth Thomas

2013-09-01

The U.S. Department of Energy’s Light Water Reactor Sustainability (LWRS) program is collaborating with a U.S. nuclear utility to bring about a systematic fleet-wide control room modernization. To facilitate this upgrade, a new distributed control system (DCS) is being introduced into the control rooms of these plants. The DCS will upgrade the legacy plant process computer and emergency response facility information system. In addition, the DCS will replace an existing analog turbine control system with a display-based system. With technology upgrades comes the opportunity to improve the overall human-system interaction between the operators and the control room. To optimize operator performance, the LWRS Control Room Modernization research team followed a human-centered approach published by the U.S. Nuclear Regulatory Commission. NUREG-0711, Rev. 3, Human Factors Engineering Program Review Model (O’Hara et al., 2012), prescribes four phases for human factors engineering. This report provides examples of the first phase, Planning and Analysis. The three elements of Planning and Analysis in NUREG-0711 that are most crucial to initiating control room upgrades are: • Operating Experience Review: Identifies opportunities for improvement in the existing system and provides lessons learned from implemented systems. • Function Analysis and Allocation: Identifies which functions at the plant may be optimally handled by the DCS vs. the operators. • Task Analysis: Identifies how tasks might be optimized for the operators. Each of these elements is covered in a separate chapter. Examples are drawn from workshops with reactor operators that were conducted at the LWRS Human System Simulation Laboratory HSSL and at the respective plants. The findings in this report represent generalized accounts of more detailed proprietary reports produced for the utility for each plant. The goal of this LWRS report is to disseminate the technique and provide examples sufficient to

Design of water feeding system for IASCC irradiation tests at JMTR

International Nuclear Information System (INIS)

Kanno, Masaru; Nabeya, Hideaki; Mori, Yuichiro

2001-12-01

In relation to the aging of light water reactors (LWRs), the irradiation assisted stress corrosion cracking (IASCC) has been regarded as a significant and urgent issue for the reliability of in-core components and materials of LWRs, and the irradiation research is now under schedule. It is essential for IASCC studies to irradiated materials under well-controlled conditions simulating LWR in-core environment. Therefore, a new water feeding system to supply high temperature water into irradiation capsules in the Japan Materials Testing Reactor (JMTR) has been designed and will be installed in near future. This report describes the specification and performance of the water feeding system that is designed to supply high temperature water to simulate BWR conditions in irradiation capsules. This design work was performed in the fiscal year 1999. (author)

Guidebook on quality control of mixed oxides and gadolinium bearing fuels for light water reactors

International Nuclear Information System (INIS)

1991-02-01

Under the coverage of an efficient quality assurance system, quality control in nuclear fuel fabrication is an essential element to assure the reliable performance of all its components in service. Incentives to increase fuel performance, by extending reactor cycles or achieving higher burnups and, in some countries to use recycled plutonium in light water reactors (LWRs) necessitated the development of new types of fuels. In the first case, due to higher uranium enrichments, a burnable neutron absorber was integrated to the fuel pellets. Gadolinia was found to form a solid solution with Uranium dioxide and, to present a burnup rate which matches fissile uranium depletion. (U,Gd)O 2 fuels which have been successfully used since the seventies, in boiling water reactors have more recently found an increased utilization, in pressurized water reactors. This amply justifies the publication of this TECDOC to encourage authorities, designers and manufacturers of these types of fuel to establish a more uniform, adapted and effective system of control, thus promoting improved materials reliability and good performance in advanced fuel for light water reactors. The Guidebook is subdivided into four chapters written by different authors. A separate abstract was prepared for each of these chapters. Refs, figs and tabs

14. Internal symposium on secular change of structural materials for nuclear energy. Secular change mechanism in light water reactor environment

International Nuclear Information System (INIS)

1993-01-01

At this symposium, lectures were given on the embrittlement by neutron irradiation of LWR pressure vessel steel, the effect that neutron irradiation exerts to austenitic stainless steel becoming sensitive, the mechanism of the occurrence and development of stress corrosion cracking in the water environment of LWRs, the effect that the water environment of LWRs exerts to fatigue life, and the environment-promoted cracking in LWR environment and its forecast. Thereafter, panel discussion was held by the lecturers. In this book, the summaries of the lectures are collected. (K.I.)

Roadmap for Nondestructive Evaluation of Reactor Pressure Vessel Research and Development by the Light Water Reactor Sustainability Program

Energy Technology Data Exchange (ETDEWEB)

Smith, Cyrus M [ORNL; Nanstad, Randy K [ORNL; Clayton, Dwight A [ORNL; Matlack, Katie [Georgia Institute of Technology; Ramuhalli, Pradeep [Pacific Northwest National Laboratory (PNNL); Light, Glenn [Southwest Research Institute, San Antonio

2012-09-01

The Department of Energy s (DOE) Light Water Reactor Sustainability (LWRS) Program is a five year effort which works to develop the fundamental scientific basis to understand, predict, and measure changes in materials and systems, structure, and components as they age in environments associated with continued long-term operations of existing commercial nuclear power reactors. This year, the Materials Aging and Degradation (MAaD) Pathway of this program has placed emphasis on emerging Non-Destructive Evaluation (NDE) methods which support these objectives. DOE funded Research and Development (R&D) on emerging NDE techniques to support commercial nuclear reactor sustainability is expected to begin next year. This summer, the MAaD Pathway invited subject matter experts to participate in a series of workshops which developed the basis for the research plan of these DOE R&D NDE activities. This document presents the results of one of these workshops which are the DOE LWRS NDE R&D Roadmap for Reactor Pressure Vessels (RPV). These workshops made a substantial effort to coordinate the DOE NDE R&D with that already underway or planned by the Electric Power Research Institute (EPRI) and the Nuclear Regulatory Commission (NRC) through their representation at these workshops.

The Consortium for Advanced Simulation of Light Water Reactors

International Nuclear Information System (INIS)

Szilard, Ronaldo; Zhang, Hongbin; Kothe, Douglas; Turinsky, Paul

2011-01-01

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is a DOE Energy Innovation Hub for modeling and simulation of nuclear reactors. It brings together an exceptionally capable team from national labs, industry and academia that will apply existing modeling and simulation capabilities and develop advanced capabilities to create a usable environment for predictive simulation of light water reactors (LWRs). This environment, designated as the Virtual Environment for Reactor Applications (VERA), will incorporate science-based models, state-of-the-art numerical methods, modern computational science and engineering practices, and uncertainty quantification (UQ) and validation against data from operating pressurized water reactors (PWRs). It will couple state-of-the-art fuel performance, neutronics, thermal-hydraulics (T-H), and structural models with existing tools for systems and safety analysis and will be designed for implementation on both today's leadership-class computers and the advanced architecture platforms now under development by the DOE. CASL focuses on a set of challenge problems such as CRUD induced power shift and localized corrosion, grid-to-rod fretting fuel failures, pellet clad interaction, fuel assembly distortion, etc. that encompass the key phenomena limiting the performance of PWRs. It is expected that much of the capability developed will be applicable to other types of reactors. CASL's mission is to develop and apply modeling and simulation capabilities to address three critical areas of performance for nuclear power plants: (1) reduce capital and operating costs per unit energy by enabling power uprates and plant lifetime extension, (2) reduce nuclear waste volume generated by enabling higher fuel burnup, and (3) enhance nuclear safety by enabling high-fidelity predictive capability for component performance.

Rate theory scenarios study on fission gas behavior of U 3 Si 2 under LOCA conditions in LWRs

Energy Technology Data Exchange (ETDEWEB)

Miao, Yinbin; Gamble, Kyle A.; Andersson, David; Mei, Zhi-Gang; Yacout, Abdellatif M.

2018-01-01

Fission gas behavior of U3Si2 under various loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs) was simulated using rate theory. A rate theory model for U3Si2 that covers both steady-state operation and power transients was developed for the GRASS-SST code based on existing research reactor/ion irradiation experimental data and theoretical predictions of density functional theory (DFT) calculations. The steady-state and LOCA condition parameters were either directly provided or inspired by BISON simulations. Due to the absence of in-pile experiment data for U3Si2's fuel performance under LWR conditions at this stage of accident tolerant fuel (ATF) development, a variety of LOCA scenarios were taken into consideration to comprehensively and conservatively evaluate the fission gas behavior of U3Si2 during a LOCA.

Physics methods for calculating light water reactor increased performances

International Nuclear Information System (INIS)

Vandenberg, C.; Charlier, A.

1988-01-01

The intensive use of light water reactors (LWRs) has induced modification of their characteristics and performances in order to improve fissile material utilization and to increase their availability and flexibility under operation. From the conceptual point of view, adequate methods must be used to calculate core characteristics, taking into account present design requirements, e.g., use of burnable poison, plutonium recycling, etc. From the operational point of view, nuclear plants that have been producing a large percentage of electricity in some countries must adapt their planning to the need of the electrical network and operate on a load-follow basis. Consequently, plant behavior must be predicted and accurately followed in order to improve the plant's capability within safety limits. The Belgonucleaire code system has been developed and extensively validated. It is an accurate, flexible, easily usable, fast-running tool for solving the problems related to LWR technology development. The methods and validation of the two computer codes LWR-WIMS and MICROLUX, which are the main components of the physics calculation system, are explained

Light Water Reactor Sustainability Research and Development Program Plan. Fiscal Year 2009-2013

International Nuclear Information System (INIS)

2009-01-01

Nuclear power has reliably and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. By the year 2030, domestic demand for electrical energy is expected to grow to levels of 16 to 36% higher than 2007 levels. At the same time, most currently operating nuclear power plants will begin reaching the end of their 60-year operating licenses. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline - even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary this year. U.S. regulators have begun considering extended operations of nuclear power plants and the research needed to support long-term operations. The Light Water Reactor Sustainability (LWRS) Research and Development (R and D) Program, developed and sponsored by the Department of Energy, is performed in close collaboration with industry R and D programs. The purpose of the LWRS R and D Program is to provide technical foundations for licensing and managing long-term, safe and economical operation of the current operating nuclear power plants. The LWRS R and D Program vision is captured in the following statements: Existing operating nuclear power plants will continue to safely provide clean and economic electricity well beyond their first license- extension period, significantly contributing to reduction of United States and global carbon emissions, enhancement of national energy security, and protection of the environment. There is a comprehensive technical basis for licensing and managing the long-term, safe, economical operation of nuclear power plants. Sustaining the existing operating U.S. fleet also will improve its

JAIF report on small- and medium-size LWRs

International Nuclear Information System (INIS)

Anon.

1983-01-01

The Ministry of International Trade and Industry has made a report on the results of the study on the utilization of small and medium size LWRs. The Japan Atomic Industrial Forum undertook this research work by the commission of the Ministry. It studied six cases of PWRs and BWRs with thermal output power ranging from 20 MWt to 700 MWt as the hypothetical design for their siting conditions, reactor capacity and the examples of utilization of the corresponding steam output. A large housing estate of 5,000 units, a local core city with population of 100,000, industrial utilization for petrochemical plants and paper and pulp plants and the combined utilization for a factory and a housing estate of 5,000 units were examined. The report said that a city with population of 100,000 would be the best object to supply heat from a nuclear reactor, and it is important to develop the nuclear reactors with 200 MWt capacity. This type is versatile in the siting conditions, and can be constructed near a city, accordingly, it offers the possibility of extensive utilization. The significance of small and medium size LWRs as the alternative energy for oil, the regional development accompanying the use of small and medium size LWRs, the technical development toward scale reduction, easy operation and versatile utilization, the economical evaluation of the utilization models and the tasks for the future are reported. (Kako, I.)

Light Water Reactor Sustainability Program: Integrated Program Plan

Energy Technology Data Exchange (ETDEWEB)

NONE

2016-02-15

proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer-than-initially-licensed lifetime. It has two facets with respect to long-term operations: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the industry to implement technology to exceed the performance of the current labor-intensive business model.

Light Water Reactor Sustainability Program: Integrated Program Plan

International Nuclear Information System (INIS)

2016-02-01

and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program's plans. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer-than-initially-licensed lifetime. It has two facets with respect to long-term operations: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the industry to implement technology to exceed the performance of the current labor-intensive business model.

Light Water Reactor Sustainability Program: Integrated Program Plan

Energy Technology Data Exchange (ETDEWEB)

None, None

2017-05-01

proliferation and terrorism. The Light Water Reactor Sustainability (LWRS) Program is the primary programmatic activity that addresses Objective 1. This document summarizes the LWRS Program’s plans. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for a longer-than-initially-licensed lifetime. It has two facets with respect to long-term operations: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the industry to implement technology to exceed the performance of the current labor-intensive business model.

Accident source terms for Light-Water Nuclear Power Plants. Final report

International Nuclear Information System (INIS)

Soffer, L.; Burson, S.B.; Ferrell, C.M.; Lee, R.Y.; Ridgely, J.N.

1995-02-01

In 1962 tile US Atomic Energy Commission published TID-14844, ''Calculation of Distance Factors for Power and Test Reactors'' which specified a release of fission products from the core to the reactor containment for a postulated accident involving ''substantial meltdown of the core''. This ''source term'', tile basis for tile NRC's Regulatory Guides 1.3 and 1.4, has been used to determine compliance with tile NRC's reactor site criteria, 10 CFR Part 100, and to evaluate other important plant performance requirements. During the past 30 years substantial additional information on fission product releases has been developed based on significant severe accident research. This document utilizes this research by providing more realistic estimates of the ''source term'' release into containment, in terms of timing, nuclide types, quantities and chemical form, given a severe core-melt accident. This revised ''source term'' is to be applied to the design of future light water reactors (LWRs). Current LWR licensees may voluntarily propose applications based upon it

Safety aspects of using gadolinium as burnable poison in pressurized water reactors

International Nuclear Information System (INIS)

Vandenberg, C.; Bonet, H.; Charlier, A.; Motte, F.

1979-01-01

Within the framework of an experimental program on the behavior of gadolinium in light water reactors (LWRs), the BR3 power plant, a small 11-MW(electric) pressurized water reactor, was operated successfully with a core containing 5% Gd 2 O 3 -UO 2 rods. The core reached an average burnup increase of 22,000 MWd/tM, corresponding to 500 effective full-power days in a single cycle. These results were used to extrapolate the consequences on safety of extending such a control policy to large LWRs. In this context, the following factors were investigated: impact on the design, reactivity control and core behavior operated with lower and more constant boric acid concentration, environmental impact, fuel handling, etc

Studies on the feasibility of the LWRs waste-thorium in-core fuel cycle in the Gas Turbine-Modular Helium Reactor

International Nuclear Information System (INIS)

Talamo, Alberto

2006-01-01

The capability to operate on LWRs waste constitutes one of the major benefits of the Gas Turbine-Modular Helium Reactor; in this paper, it has been evaluated the possibility to incinerate the LWRs waste and to simultaneously breed fissile 233 U by fertile thorium. Since a mixture of pure 239 Pu-thorium has shown a quite poor neutron economy, the LWRs waste-thorium fuel performance has been also tested when plutonium and thorium are allocated in different TRISO particles. More precisely, when fissile and fertile actinides share the same TRISO kernel, the resonance at 0.29eV of the fission and capture microscopic cross sections of 239 Pu diminishes also the absorption rate of fertile 232 Th and thus it degrades the breeding process. Consequently, in the present studies, two different types of fuel have been utilized: the Driver Fuel, made of LWRs waste, and the Transmutation Fuel, made of fertile thorium. Since, in the thermal neutron energy range, the microscopic capture cross section of 232 Th is about 80-100 times smaller than the fission one of 239 Pu, setting thorium in particles with a large kernel and LWRs waste in particles with a small one makes the volume integrated reaction rates better equilibrated. At the light of the above consideration, which drives to load as much thorium as possible, for the Transmutation Fuel they have been selected the JAERI TRISO particles packed 40%; whereas, for the Driver Fuel they have been tested different packing fractions and kernel radii. Since no configuration allowed the reactor to work, the above procedure has been repeated when fertile particles are packed 20%; the latter choice permits over one year of operation, but the build up of 233 U represents only a small fraction of the depleted 239 Pu. Finally, the previous configuration has been also investigated when the fertile and fissile fuels share the same kernel or when the fertile fuel axially alternates with the fissile one. (author)

Proceedings of the Second OECD/NEA Organisation Meeting on Increased Accident Tolerance of Fuels for LWRs

International Nuclear Information System (INIS)

Massara, Simone; ); Bragg-Sitton, Shannon; Braase, Lori; Merrill, Brad; Teague, Melissa; Stanek, Chris R.; Montgomery, Robert H.; Ott, Larry J.; Robb, Kevin; Snead, Lance; Farmer, Mitch; Billone, Michael C.; Todosow, Michael; Brown, Nicholas; Brachet, J.C.; Le Flem, M.; Sauder, C.; Idarraga-Trujillo, I.; Michaux, A.; Lorrette, C.; Le Saux, M.; Blanpain, P.; Park, Jeong-Yong; Yang, Jae-Ho; Kim, Weon-Ju; Koo, Yang-Hyun; Liu, T.; Hallstadius, Lars; Lahoda, Ed; Waeckel, N.; Bonnet, J.M.; Vitanza, Carlo; Ohta, Hirokazu; Ogata, Takanari; Nakamura, Kinya; Dyck, Gary; Inozemtsev, Victor; )

2013-01-01

Under the guidance of the OECD-NEA Nuclear Science Committee, the expert group acts as a forum for scientific and technical information exchange on advanced light water reactor (LWR) fuels with enhanced accident tolerance. The expert group focusses on the fundamental properties and behaviour under normal operations and accident conditions for advanced core materials and components (fuels, cladding, control rods, etc.). The materials considered are applicable to Gen II and Gen III Light Water Reactors, as well as Gen III+ reactors under construction. The objective of the expert group is to define and coordinate a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with enhanced accident tolerance compared to currently used zircaloy/UO 2 fuel systems, as well as other non-fuel core components with important roles in LWR performance under accident conditions. This document brings together the available presentations (slides) given at the 2. Meeting on Increased Accident Tolerance of Fuels for LWRs. Content: 1 - Overview of the exchanges after the December-2012 Workshop through the discussion forum established at the OECD-NEA (S. Massara, NEA); 2 - Metrics Development for Enhanced Accident Tolerant LWR Fuels (S. Bragg-Sitton, INL); 3 - Candidate ATF Clad Technologies and Key Feasibility Issues (L. Snead, ORNL); 4 - CEA studies on nuclear fuel claddings for LWRs enhanced accident tolerant fuel. Some recent results, pending issues and prospects (J.C. Brachet, CEA); 5 - Current status on the accident tolerant fuel development in the Republic of Korea (J.Y. Park, J.H. Chang, KAERI); 6 - The current status of fuel R and D in the P.R. of China (T. Liu, CGN). Session 2: Key elements for a work programme on ATF: 7 - Beneficial characteristics of ATF (metrics) (L. Hallstadius, Westinghouse); 8 - Reactor types of interest (applicability) (L. Ott, ORNL); 9 - Impact on normal operations

Light Water Reactor Sustainability Research and Development Program Plan -- Fiscal Year 2009–2013

Energy Technology Data Exchange (ETDEWEB)

Idaho National Laboratory

2009-12-01

Nuclear power has reliably and economically contributed almost 20% of electrical generation in the United States over the past two decades. It remains the single largest contributor (more than 70%) of non-greenhouse-gas-emitting electric power generation in the United States. By the year 2030, domestic demand for electrical energy is expected to grow to levels of 16 to 36% higher than 2007 levels. At the same time, most currently operating nuclear power plants will begin reaching the end of their 60-year operating licenses. If current operating nuclear power plants do not operate beyond 60 years, the total fraction of generated electrical energy from nuclear power will begin to decline—even with the expected addition of new nuclear generating capacity. The oldest commercial plants in the United States reached their 40th anniversary this year. U.S. regulators have begun considering extended operations of nuclear power plants and the research needed to support long-term operations. The Light Water Reactor Sustainability (LWRS) Research and Development (R&D) Program, developed and sponsored by the Department of Energy, is performed in close collaboration with industry R&D programs. The purpose of the LWRS R&D Program is to provide technical foundations for licensing and managing long-term, safe and economical operation of the current operating nuclear power plants. The LWRS R&D Program vision is captured in the following statements: Existing operating nuclear power plants will continue to safely provide clean and economic electricity well beyond their first license- extension period, significantly contributing to reduction of United States and global carbon emissions, enhancement of national energy security, and protection of the environment. There is a comprehensive technical basis for licensing and managing the long-term, safe, economical operation of nuclear power plants. Sustaining the existing operating U.S. fleet also will improve its international engagement

The nuclear reactor strategy between fast breeder reactors and advanced pressurized water reactors

International Nuclear Information System (INIS)

Seifritz, W.

1983-01-01

A nuclear reactor strategy between fast breeder reactors (FBRs) and advanced pressurized water reactors (APWRs) is being studied. The principal idea of this strategy is that the discharged plutonium from light water reactors (LWRs) provides the inventories of the FBRs and the high-converter APWRs, whereby the LWRs are installed according to the derivative of a logistical S curve. Special emphasis is given to the dynamics of reaching an asymptotic symbiosis between FBRs and APWRs. The main conclusion is that if a symbiotic APWR-FBR family with an asymptotic total power level in the terawatt range is to exist in about half a century from now, we need a large number of FBRs already in an early phase

Radionuclide distribution in LWR [light-water reactor] spent fuel

International Nuclear Information System (INIS)

Guenther, R.J.; Blahnik, D.E.; Thomas, L.E.; Baldwin, D.L.; Mendel, J.E.

1990-09-01

The Materials Characterization Center (MCC) at Pacific Northwest Laboratory (PNL) provides well-characterized spent fuel from light-water reactors (LWRs) for use in laboratory tests relevant to nuclear waste disposal in the proposed Yucca Mountain repository. Interpretation of results from tests on spent fuel oxidation, dissolution, and cladding degradation requires information on the inventory and distribution of radionuclides in the initial test materials. The MCC is obtaining this information from examinations of Approved Testing Materials (ATMs), which include spent fuel with burnups from 17 to 50 MWd/kgM and fission gas releases (FGR) from 0.2 to 18%. The concentration and distribution of activation products and the release of volatile fission products to the pellet-cladding gap and rod plenum are of particular interest because these characteristics are not well understood. This paper summarizes results that help define the 14 C inventory and distribution in cladding, the ''gap and grain boundary'' inventory of radionuclides in fuels with different FGRs, and the structure and radionuclide inventory of the fuel rim region within a few hundred micrometers from the fuel edge. 6 refs., 5 figs., 1 tab

Potential for fuel melting and cladding thermal failure during a PCM event in LWRs

International Nuclear Information System (INIS)

El-Genk, M.S.; Croucher, D.W.

1979-01-01

The primary concern in nuclear reactor safety is to ensure that no conceivable accident, whether initiated by a failure of the reactor system or by incorrect operation, will lead to a dangerous release of radiation to the environment. A number of hypothesized off-normal power or cooling conditions, generally termed as power-cooling-mismatch (PCM) accidents, are considered in the safety analysis of light water reactors (LWRs). During a PCM accident, film boiling may occur at the cladding surface and cause a rapid temperature increase in the fuel and the cladding, perhaps producing embrittlement of the zircaloy cladding by oxidation. Molten fuel may be produced at the center of the pellets, extrude radially through open cracks in the outer, unmelted portion of the pellet and relocate in the fuel-cladding gap. If the amount of extruded molten fuel is sufficient to establish contact with the cladding, which is at a high temperature during film boiling, the zircaloy cladding may melt. The present work assesses the potential for central fuel melting and thermal failure of the zircaloy cladding due to melting upon being contacted by extruded molten UO 2 -fuel during a PCM event

Light Water Reactor Sustainability Program: Digital Technology Business Case Methodology Guide

Energy Technology Data Exchange (ETDEWEB)

Thomas, Ken [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lawrie, Sean [ScottMadden, Inc., Raleigh, NC (United States); Hart, Adam [ScottMadden, Inc., Raleigh, NC (United States); Vlahoplus, Chris [ScottMadden, Inc., Raleigh, NC (United States)

2014-09-01

The Department of Energy’s (DOE’s) Light Water Reactor Sustainability Program aims to develop and deploy technologies that will make the existing U.S. nuclear fleet more efficient and competitive. The program has developed a standard methodology for determining the impact of new technologies in order to assist nuclear power plant (NPP) operators in building sound business cases. The Advanced Instrumentation, Information, and Control (II&C) Systems Technologies Pathway is part of the DOE’s Light Water Reactor Sustainability (LWRS) Program. It conducts targeted research and development (R&D) to address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals: (1) to ensure that legacy analog II&C systems are not life-limiting issues for the LWR fleet and (2) to implement digital II&C technology in a manner that enables broad innovation and business improvement in the NPP operating model. Resolving long-term operational concerns with the II&C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation’s energy and environmental security. The II&C Pathway is conducting a series of pilot projects that enable the development and deployment of new II&C technologies in existing nuclear plants. Through the LWRS program, individual utilities and plants are able to participate in these projects or otherwise leverage the results of projects conducted at demonstration plants. Performance advantages of the new pilot project technologies are widely acknowledged, but it has proven difficult for utilities to derive business cases for justifying investment in these new capabilities. Lack of a business case is often cited by utilities as a barrier to pursuing wide-scale application of digital technologies to nuclear plant work activities. The decision to move forward with funding usually hinges on

Light water reactor (LWR) innovation needs in the United States: The Massachusetts Institute of Technology LWR innovation project

International Nuclear Information System (INIS)

Golay, M.W.

1988-01-01

A major effort under way within the Massachusetts Institute of Technology (MIT) Engineering School is focused on the contributions that technology innovation can make in revitalizing nuclear power in the United States. A principal component of this effort is a project to improve the designs of the next generation of light water reactors (LWRs) with emphasis on achieving improved capacity factors and safety, and reducing the construction duration. The motivation for this overall effort is to prevent the nuclear option from being unnecessarily lost by being available only in uneconomic configurations. In considering how to advance this effort, the authors focused on refining the designs of new reactors because this is the area where the greatest opportunities for improvements exist

Proceedings of the Start-up Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs, 28-29 April 2014, OECD-NEA HQ

International Nuclear Information System (INIS)

Kurata, Masaki; Bragg-Sitton, Shannon; Pasamehmetoglu, K.; Sowder, Andrew; Koo, Yang-Hyun; Yang, Jae-Ho; Kim, Hyun-Gil; Zhou, Y.; Forgeron, T.; Guedeney, Ph.; Brachet, J.C.; Michaux, A.; Chauvin, Nathalie; Waeckel, N.; Ambard, A.; Blanpain, P.; Bischoff, J.; Zvonarev, Yu.; Verwerft, M.; Weber, M.; Lambrinou, K.; Koonen, E.; Van Dyck, S.; PETIT, Marc; Cornet, Stephanie; ); YAMAJI, Akifumi; ); Inozemtsev, V.; )

2014-04-01

Under the guidance of the OECD-NEA Nuclear Science Committee, the expert group acts as a forum for scientific and technical information exchange on advanced light water reactor (LWR) fuels with enhanced accident tolerance. The expert group focusses on the fundamental properties and behaviour under normal operations and accident conditions for advanced core materials and components (fuels, cladding, control rods, etc.). The materials considered are applicable to Gen II and Gen III Light Water Reactors, as well as Gen III+ reactors under construction. The objective of the expert group is to define and coordinate a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with enhanced accident tolerance compared to currently used zircaloy/UO 2 fuel systems, as well as other non-fuel core components with important roles in LWR performance under accident conditions. This document brings together the available presentations (slides) given at the Start-up Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs. Content: 1 - Final Agenda; 2 - Draft mandate of EGATFL: Discussion of Scope and Objectives (K. Pasamehmetoglu, INL); 3 - Technical updates since the 2. meeting on ATF (28-29 October 2013): - Overview on ATF R and D in Japan (M. Kurata, JAEA); - Update on Development of Enhanced Accident Tolerant Fuel for Light Water Reactors in the United States (S. Bragg-Sitton, INL); - EPRI Update Since the 2. OECD/NEA Meeting on ATF - 28-29 October 2013 (A. Sowder, EPRI); - Accident Tolerant Fuel (ATF) Development: KAERI's R and D Status (Y.H. Koo, KAERI); - Accident Tolerant Fuel Research Activities in China General Nuclear Power Corporation - CGN (Y. Zhou, CGN); - ATF R and D Status and Perspectives (Th. Forgeron, CEA); - Proposals of NRC 'Kurchatov Institute' on Contributions to Collaborative Framework on ATF Activity (Y. Zvonarev, NRC KI); - Input to the

MOX fuel fabrication and utilisation in LWRs worldwide

International Nuclear Information System (INIS)

Provost, J.-L.; Schrader, M.; Nomura, S.

2000-01-01

Early in the development of the nuclear programme, a large part of the countries using nuclear energy has studied the reprocessing and recycling option in order to develop a safe conditioning of fission products and to recycle fissile materials in reactors. In the sixties, the feasibility of recycling plutonium in LWRs has been successfully demonstrated by several experimentations of MOX rod irradiations in different countries. Based on the background of the MOX behaviour collected during the seventies and on the results of the important MOX experimentation program implemented during this period, a large part of the European utilities decided at the beginning of the eighties to use MOX fuel in LWRs on an industrial scale. The main goals of the utilities were to use as a fuel an available fissile material and to control the stockpile of separated plutonium. Today, the understanding of the behaviour of plutonium fuel has grown significantly since the launch of the first R and D programmes on LWR and FR MOX fuels. Plutonium oxide physical and neutron behaviour is well known, its modelling is now available as well as experimentally validated. Up to now, more than 750 tHM MOX fuel (more than 2000 FAs) have been loaded in 29 PWRs and in 2 BWRs in Europe, corresponding to the recycling of about 35 t of plutonium. Reprocessing/recycling technology has reached maturity in the main nuclear industry countries. Spent fuel reprocessing and recycling of the separated fissile materials remains the main option for the back-end cycle. Today, the operation of MOX-recycling LWRs is considered satisfactory. Experience feedback shows that, in global terms, MOX cores behaviour is equivalent to that of UO 2 cores in terms of operation and safety. (author)

Report from the Light Water Reactor Sustainability Workshop on On-Line Monitoring Technologies

International Nuclear Information System (INIS)

Baldwin, Thomas; Tawfik, Magdy; Bond, Leonard

2010-01-01

In support of expanding the use of nuclear power, interest is growing in methods of determining the feasibility of longer term operation for the U.S. fleet of nuclear power plants, particularly operation beyond 60 years. To help establish the scientific and technical basis for such longer term operation, the DOE-NE has established a research and development (R and D) objective. This objective seeks to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of current reactors. The Light Water Reactor Sustainability (LWRS) Program, which addresses the needs of this objective, is being developed in collaboration with industry R and D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of nuclear power plants. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. In moving to identify priorities and plan activities, the Light Water Reactor Sustainability Workshop on On-Line Monitoring (OLM) Technologies was held June 10-12, 2010, in Seattle, Washington. The workshop was run to enable industry stakeholders and researchers to identify the nuclear industry needs in the areas of future OLM technologies and corresponding technology gaps and research capabilities. It also sought to identify approaches for collaboration that would be able to bridge or fill the technology gaps. This report is the meeting proceedings, documenting the presentations and discussions of the workshop and is intended to serve as a basis for a plan which is under development that will enable the I and C research pathway to achieve its goals. Benefits to the nuclear industry accruing from On Line Monitoring Technology cannot be ignored. Information gathered thus far has contributed significantly to the Department of Energy's Light Water Reactor Sustainability Program. DOE

Report from the Light Water Reactor Sustainability Workshop on On-Line Monitoring Technologies

Energy Technology Data Exchange (ETDEWEB)

Thomas Baldwin; Magdy Tawfik; Leonard Bond

2010-06-01

In support of expanding the use of nuclear power, interest is growing in methods of determining the feasibility of longer term operation for the U.S. fleet of nuclear power plants, particularly operation beyond 60 years. To help establish the scientific and technical basis for such longer term operation, the DOE-NE has established a research and development (R&D) objective. This objective seeks to develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of current reactors. The Light Water Reactor Sustainability (LWRS) Program, which addresses the needs of this objective, is being developed in collaboration with industry R&D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of nuclear power plants. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. In moving to identify priorities and plan activities, the Light Water Reactor Sustainability Workshop on On-Line Monitoring (OLM) Technologies was held June 10–12, 2010, in Seattle, Washington. The workshop was run to enable industry stakeholders and researchers to identify the nuclear industry needs in the areas of future OLM technologies and corresponding technology gaps and research capabilities. It also sought to identify approaches for collaboration that would be able to bridge or fill the technology gaps. This report is the meeting proceedings, documenting the presentations and discussions of the workshop and is intended to serve as a basis for a plan which is under development that will enable the I&C research pathway to achieve its goals. Benefits to the nuclear industry accruing from On Line Monitoring Technology cannot be ignored. Information gathered thus far has contributed significantly to the Department of Energy’s Light Water Reactor Sustainability Program. DOE has

Propertis of solidified radioactive wastes from commercial LWRs

International Nuclear Information System (INIS)

Neilson, R.M. Jr.; Colombo, P.

1978-01-01

A study has been performed to characterize the properties of solidified radioactive wastes generated in the liquid radwaste treatment systems at LWRs. The properties which have been studied are those which are pertinent in defining the relative potential for the release of radionuclides to the environment as well as others relating to the evaluation of various solidification agents on an economic and feasibility basis. The use of standard testing procedures in measuring these properties allows an intercomparison of respective properties between various types of solidified waste forms. The leachability, mechanical properties, thermal stability, radiation stability, and thermal properties of hydraulic cement, ureaformaldehyde, bitumen, and addition type polymer waste forms have been measured. In addition, the chemical sensitivity, volumetric efficiency and radiation shielding characteristics of these waste forms have been studied. Emphasis in this paper is placed on the results of studies concerning chemical compatibility of solidification agents with specific waste streams, volumetric efficiency, free standing water, and leachability

Fast reactor cooled by supercritical light water

Energy Technology Data Exchange (ETDEWEB)

Ishiwatari, Yuki; Mukouhara, Tami; Koshizuka, Seiichi; Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Lab., Tokai, Ibaraki (Japan)

2001-09-01

This report introduces the result of a feasibility study of a fast reactor cooled by supercritical light water (SCFR) with once-through cooling system. It is characterized by (1) no need of steam separator, recirculation system, or steam generator, (2) 1/7 of core flow rate compared with BWR or PWR, (3) high temperature and high pressure permits small turbine and high efficiency exceeding 44%, (4) structure and operation of major components are already experienced by LWRs or thermal power plants. Modification such as reducing blanket fuels and increasing seed fuels are made to achieve highly economic utilization of Pu and high power (2 GWe). The following restrictions were satisfied. (1) Maximum linear heat rate 39 kW/m, (2) Maximum surface temperature of Inconel cladding 620degC, (3) Negative void reactivity coefficient, (4) Fast neutron irradiation rate at the inner surface of pressure vessel less than 2.0x10{sup 19} n/cm{sup 2}. Thus the high power density of 167 MW/m{sup 3} including blanket is thought to contributes economy. The high conversion is attained to be 0.99 Pu fission residual rate by the outer radius of fuel rod of 0.88 mm. The breeding of 1.034 by Pu fission residual rate can be achieved by using briquette (tube-in-shell) type fuel structure. (K. Tsuchihashi)

Reducing occupational radiation exposures at LWRs

International Nuclear Information System (INIS)

Lattanzi, D.; Neri, C.; Papa, C.; Paribelli, S.

1980-01-01

The occupational radiation doses received by nuclear power plant personnel during a period of several years of operation are briefly reviewed. Comparisons are made between the data for BWRs and PWRs in order to identify the more ''critical'' reactor type, from a radiological poin; of view. Attention is also devoted to GCRs. Furthermore the areas which contribute most to personnel doses are considered and briefly reviewed. The main steps to be taken in order to reduce occupational radiation exposures at LWRs are discussed. (H.K.)

Environmentally assisted cracking of light-water reactor materials

International Nuclear Information System (INIS)

Chopra, O.K.; Chung, H.M.; Kassner, T.F.; Shack, W.J.

1996-02-01

Environmentally assisted cracking (EAC) of lightwater reactor (LWR) materials has affected nuclear reactors from the very introduction of the technology. Corrosion problems have afflicted steam generators from the very introduction of pressurized water reactor (PWR) technology. Shippingport, the first commercial PWR operated in the United States, developed leaking cracks in two Type 304 stainless steel (SS) steam generator tubes as early as 1957, after only 150 h of operation. Stress corrosion cracks were observed in the heat-affected zones of welds in austenitic SS piping and associated components in boiling-water reactors (BRWs) as early as 1965. The degradation of steam generator tubing in PWRs and the stress corrosion cracking (SCC) of austenitic SS piping in BWRs have been the most visible and most expensive examples of EAC in LWRs, and the repair and replacement of steam generators and recirculation piping has cost hundreds of millions of dollars. However, other problems associated with the effects of the environment on reactor structures and components am important concerns in operating plants and for extended reactor lifetimes. Cast duplex austenitic-ferritic SSs are used extensively in the nuclear industry to fabricate pump casings and valve bodies for LWRs and primary coolant piping in many PWRs. Embrittlement of the ferrite phase in cast duplex SS may occur after 10 to 20 years at reactor operating temperatures, which could influence the mechanical response and integrity of pressure boundary components during high strain-rate loading (e.g., seismic events). The problem is of most concern in PWRs where slightly higher temperatures are typical and cast SS piping is widely used

Safeguards and security requirements for weapons plutonium disposition in light water reactors

International Nuclear Information System (INIS)

Thomas, L.L.; Strait, R.S.

1994-10-01

This paper explores the issues surrounding the safeguarding of the plutonium disposition process in support of the United States nuclear weapons dismantlement program. It focuses on the disposition of the plutonium by burning mixed oxide fuel in light water reactors (LWR) and addresses physical protection, material control and accountability, personnel security and international safeguards. The S and S system needs to meet the requirements of the DOE Orders, NRC Regulations and international safeguards agreements. Experience has shown that incorporating S and S measures into early facility designs and integrating them into operations provides S and S that is more effective, more economical, and less intrusive. The plutonium disposition safeguards requirements with which the US has the least experience are the implementation of international safeguards on plutonium metal; the large scale commercialization of the mixed oxide fuel fabrication; and the transportation to and loading in the LWRs of fresh mixed oxide fuel. It is in these areas where the effort needs to be concentrated if the US is to develop safeguards and security systems that are effective and efficient

Assessment of Possible Cycle Lengths for Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

International Nuclear Information System (INIS)

Sen, R. Sonat; Pope, Michael A.; Ougouag, Abderrafi M.; Pasamehmetoglu, Kemal O.

2012-01-01

The tri-isotropic (TRISO) fuel developed for High Temperature reactors is known for its extraordinary fission product retention capabilities (1). Recently, the possibility of extending the use of TRISO particle fuel to Light Water Reactor (LWR) technology, and perhaps other reactor concepts, has received significant attention (2). The Deep Burn project (3) currently focuses on once-through burning of transuranic fissile and fissionable isotopes (TRU) in LWRs. The fuel form for this purpose is called Fully-Ceramic Micro-encapsulated (FCM) fuel, a concept that borrows the TRISO fuel particle design from high temperature reactor technology, but uses SiC as a matrix material rather than graphite. In addition, FCM fuel may also use a cladding made of a variety of possible material, again including SiC as an admissible choice. The FCM fuel used in the Deep Burn (DB) project showed promising results in terms of fission product retention at high burnup values and during high-temperature transients. In the case of DB applications, the fuel loading within a TRISO particle is constituted entirely of fissile or fissionable isotopes. Consequently, the fuel was shown to be capable of achieving reasonable burnup levels and cycle lengths, especially in the case of mixed cores (with coexisting DB and regular LWR UO2 fuels). In contrast, as shown below, the use of UO2-only FCM fuel in a LWR results in considerably shorter cycle length when compared to current-generation ordinary LWR designs. Indeed, the constraint of limited space availability for heavy metal loading within the TRISO particles of FCM fuel and the constraint of low (i.e., below 20 w/0) 235U enrichment combine to result in shorter cycle lengths compared to ordinary LWRs if typical LWR power densities are also assumed and if typical TRISO particle dimensions and UO2 kernels are specified. The primary focus of this summary is on using TRISO particles with up to 20 w/0 enriched uranium kernels loaded in Pressurized Water

OECD - HRP Summer School on Light Water Reactor Structural Materials. August 26th - 30th, 2002

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-07-01

In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on Light Water Reactor Structural Materials in the period August 26 - 30, 2002. The summer school was primarily intended for people who wanted to become acquainted with materials-related subjects and issues without being experts. It is especially hoped that the summer school served to transfer knowledge to the ''young generation'' in the field of nuclear. Experts from Halden Project member organisations were solicited for the following programme: (1) Overview of The Nuclear Community and Current Issues, (2) Regulatory Framework for Ensuring Structural Integrity, (3) Non-Destructive Testing for Detection of Cracks, (4) Part I - Basics of Radiation and Radiation Damage, (5) Part II - Radiation Effects on Reactor Internal Materials, (6) Water Chemistry and Radiolysis Effects in LWRs, (7) PWR and Fast Breeder Reactor Internals, (8) PWR and Fast Breeder Reactor Internals, (9) Secondary Side Corrosion Cracking of PWR Steam Generator Tubes, (10) BWR Materials and Their Interaction with the Environment, (11) Radiation Damage in Reactor Pressure Vessels.

OECD - HRP Summer School on Light Water Reactor Structural Materials. August 26th - 30th, 2002

International Nuclear Information System (INIS)

2002-01-01

In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on Light Water Reactor Structural Materials in the period August 26 - 30, 2002. The summer school was primarily intended for people who wanted to become acquainted with materials-related subjects and issues without being experts. It is especially hoped that the summer school served to transfer knowledge to the ''young generation'' in the field of nuclear. Experts from Halden Project member organisations were solicited for the following programme: (1) Overview of The Nuclear Community and Current Issues, (2) Regulatory Framework for Ensuring Structural Integrity, (3) Non-Destructive Testing for Detection of Cracks, (4) Part I - Basics of Radiation and Radiation Damage, (5) Part II - Radiation Effects on Reactor Internal Materials, (6) Water Chemistry and Radiolysis Effects in LWRs, (7) PWR and Fast Breeder Reactor Internals, (8) PWR and Fast Breeder Reactor Internals, (9) Secondary Side Corrosion Cracking of PWR Steam Generator Tubes, (10) BWR Materials and Their Interaction with the Environment, (11) Radiation Damage in Reactor Pressure Vessels

Proceedings of the Third Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs, 3-5 March 2015, OECD-NEA HQ

International Nuclear Information System (INIS)

Bischoff, Jeremy; Gandrille, Pascal; Forgeron, Thierry; Brachet, Jean-Christophe; Lorrette, Christophe; Valot, C.; Freyss, M.; Braun, J.; Sauder, C.; Moatti, Marie; Waeckel, Nicolas; Ambard, Antoine; Pasamehmetoglu, Kemal; Johnston, Emma; Bragg-Sitton, Shannon M.; Kurata, M.; Hallstadius, Lars; Ohta, H.; Ogata, T.; Besmann, T.; Chauvin, Nathalie; Cornet, Stephanie; Massara, S.; Kohyama, Akira; Kishimoto, Hirotatsu; Park, Joon Soo; Nakazato, Naofumi; Hayasaka, Daisuke; Asakura, Yuuki; Kanda, Chisato; Kohyama, Akira; Terrani, Kurt; Katoh, Yutai; Yamamoto, Yukinori; Field, Kevin; Snead, Lance; Hu, Xunxiang; Dryepondt, Sebastien; Unocic, Kinga A.; Hoelzer, David T.; Pint, Bruce A.; Besmann, T.; Steinbrueck, M.; Grosse, M.; Jianu, A.; Weisenburger, A.; Avincola, V.; Ahmad, S.; Tang, C.; Heuser, Brent J.; Sickafus, Kurt; Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jung-Hwan; Park, Jeong-Yong; Koo, Yang-Hyun; Lee, B.O.; Van Nieuwenhove, Rudi; Kim, Young; Rebak, Raul; Dolly, Evan; Dolley, E.J.; Rebak, R.B.; Maloy, Stu; Yang, Jae-Ho; Kim, Dong-Joo; Kim, Keon-Sik; Koo, Yang-Hyun; Lee, Won Jae; Tulenko, James S.; Puide, Mattias; Liu, T.; Gueneau, C.; Gosse, S.; Dupin, N.; Barber, D.; Corcoran, E.; Dumas, J.C.; Hania, R.; Kaye, M.; Turchi, P.

2015-03-01

Under the guidance of the OECD-NEA Nuclear Science Committee, the expert group acts as a forum for scientific and technical information exchange on advanced light water reactor (LWR) fuels with enhanced accident tolerance. The expert group focusses on the fundamental properties and behaviour under normal operations and accident conditions for advanced core materials and components (fuels, cladding, control rods, etc.). The materials considered are applicable to Gen II and Gen III Light Water Reactors, as well as Gen III+ reactors under construction. The objective of the expert group is to define and coordinate a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with enhanced accident tolerance compared to currently used zircaloy/UO 2 fuel systems, as well as other non-fuel core components with important roles in LWR performance under accident conditions. This document brings together the available presentations (slides) given at the Third Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs. Content: 1 - Task Force 1 (Systems assessment) meeting, 3-4 March 2015: - French evaluation of ATF Concepts (J. Bischoff, AREVA); - Technology Readiness Levels - TRL - for Fuels (K. Pasamehmetoglu, INL); - TRL-definition for advanced fuel concept applied for commercial LWRs in Japan (M. Kurata, JAEA); - Application of TRLs in NNL (E. Johnston, NNL); - Technology Readiness Levels for Advanced Nuclear Fuel and Materials (S. Bragg-Sitton, INL); 1a - Definition of the illustrative scenarios: - AREVA's proposal concerning scenario for Accident Tolerant Fuel studies (P. Gandrille); - A Simplified Accident Scenario (L. Hallstadius); - Accident Scenarios for ATF Performance Evaluation of BWR and PWR in Japan (H. Ohta, CRIEPI); 1b - Related NEA activities: - Working Party on Multi-scale Modelling of Fuels and Structural Materials for Nuclear Systems - WPMM, Expert

Reduction of releases of radioactive effluents from light-water-power-reactors in Japan

International Nuclear Information System (INIS)

Yoshida, Y.; Itakura, T.; Kanai, T.

1977-01-01

Japan Atomic Energy Commission established the dose objectives to the population around the light-water-reactors in May, 1975, based on the ''ALAP'' concept. These values are respectively, 5 mrems per year for total body and 15 mrems per year for thyroid of an individual in the critical group in the environs, due to both gaseous and liquid effluents from LWRs in one site. The present paper describes the implications of the dose objective values, control measures which have been adopted to reduce releases of radioactive materials and related technical developments in Japan. The main control measures for reduction of radioactive gaseous effluents are an installation of a charcoal gas holdup system for decay of noble gases and a supply of clean steam for the gland seal of a turbine in BWR, and a storage tank system allowing decay of noble gases in PWR. For liquid effluents are taken measures to re-use them as the primary coolant. Consequently, the amounts of radioactivity released to the environment from any LWR during normal operation have been maintained under the level to meet the above dose objective values. For research reactors, reduction of release of effluents has also been carried out in a similar way to LWRs. In order to establish the techniques applicable for further reduction, studies are being made on the control measures to reduce leakage of radioiodine, an apparatus for removal of krypton, the treatment of laundry waste and measures to remove the crud in the primary coolant. Presentation is also made on the energy-integrated gas monitor for gaseous effluent and systems of measuring γ dose from radioactive cloud descriminating from natural background, which have been developed for effective monitoring thus reduced environmental dose

Large-scale experiments on aerosol behavior in light water reactor containments

International Nuclear Information System (INIS)

Schock, W.; Bunz, H.; Adams, R.E.; Tobias, M.L.; Rahn, F.J.

1988-01-01

Recently, three large-scale experimental programs were carried out dealing with the behavior of aerosols during core-melt accidents in light water reactors (LWRs). In the Nuclear Safety Pilot Plant (NSPP) program, the principal behaviors of different insoluble aerosols and of mixed aerosols were measured in dry air atmospheres and in condensing steam-air atmospheres contained in a 38-m/sup 3/ steel vessel. The Demonstration of Nuclear Aerosol Behavior (DEMONA) program used a 640-m/sup 3/ concrete containment model to simulate typical accident sequence conditions, and measured the behavior of different insoluble aerosols and mixed aerosols in condensing and transient atmospheric conditions. Part of the LWR Aerosol Containment Experiments (LACE) program was also devoted to aerosol behavior in containment; and 852-m/sup 3/ steel vessel was used, and the aerosols were composed of mixtures of insoluble and soluble species. The results of these experiments provide a suitable data base for validation of aerosol behavior codes. Fundamental insight into details of aerosol behavior in condensing environments has been gained through the results of the NSPP tests. Code comparisons have been and are being performed in the DEMONA and LACE experiments

Production and quality control of fuel cladding tubes for LWRs

International Nuclear Information System (INIS)

Matsuda, Katsuhiko; Hagi, Shigeki; Anada, Hiroyuki; Abe, Hideaki; Hyodo, Shigetoshi

1994-01-01

This paper reviews the recent fabrication technology and corrosion resistance study of fuel cladding tubes for LWRs conducted by Sumitomo Metal Industries Ltd. started the research on zircaloy in 1957. In 1980, the factory exclusively for the production of cladding tubes was founded, and the mass production system on full scale was established. Thereafter, the various improvement of the production technology, the development of new products, and the heightening of the performance mainly on the corrosion resistance have been tested and studied. Recently, the works in the production processes were almost automated, and the installation of the production lines advanced, and the stabilization of product quality and the rationalization of costs are promoted. Moreover, the development of the zircaloy cladding tubes having high corrosion resistance has been advanced to cope with the long term cycle operation of LWRs hereafter. The features of zircaloy cladding tubes, the manufacturing processes, the improvement of the manufacturing technology, the improvement of the corrosion resistance and so on are reported. (K.I.)

Report from the Light Water Reactor Sustainability Workshop on Advanced Instrumentation, Information, and Control Systems and Human-System Interface Technologies

International Nuclear Information System (INIS)

Hallbert, Bruce P.; Persensky, J.J.; Smidts, Carol; Aldemir, Tunc; Naser, Joseph

2009-01-01

The Light Water Reactor Sustainability (LWRS) Program is a research and development (R and D) program sponsored by the U.S. Department of Energy (DOE). The program is operated in close collaboration with industry R and D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of Nuclear Power Plants that are currently in operation. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. Advanced instruments and control (I and C) technologies are needed to support the safe and reliable production of power from nuclear energy systems during sustained periods of operation up to and beyond their expected licensed lifetime. This requires that new capabilities to achieve process control be developed and eventually implemented in existing nuclear assets. It also requires that approaches be developed and proven to achieve sustainability of I and C systems throughout the period of extended operation. The strategic objective of the LWRS Program Advanced Instrumentation, Information, and Control Systems Technology R and D pathway is to establish a technical basis for new technologies needed to achieve safety and reliability of operating nuclear assets and to implement new technologies in nuclear energy systems. This will be achieved by carrying out a program of R and D to develop scientific knowledge in the areas of: (1) Sensors, diagnostics, and prognostics to support characterization and prediction of the effects of aging and degradation phenomena effects on critical systems, structures, and components (SSCs); (2) Online monitoring of SSCs and active components, generation of information, and methods to analyze and employ online monitoring information; (3) New methods for visualization, integration, and information use to enhance state awareness and leverage expertise to achieve safer, more readily available

Report from the Light Water Reactor Sustainability Workshop on Advanced Instrumentation, Information, and Control Systems and Human-System Interface Technologies

Energy Technology Data Exchange (ETDEWEB)

Bruce P. Hallbert; J. J. Persensky; Carol Smidts; Tunc Aldemir; Joseph Naser

2009-08-01

The Light Water Reactor Sustainability (LWRS) Program is a research and development (R&D) program sponsored by the U.S. Department of Energy (DOE). The program is operated in close collaboration with industry R&D programs to provide the technical foundations for licensing and managing the long-term, safe, and economical operation of Nuclear Power Plants that are currently in operation. The LWRS Program focus is on longer-term and higher-risk/reward research that contributes to the national policy objectives of energy and environmental security. Advanced instruments and control (I&C) technologies are needed to support the safe and reliable production of power from nuclear energy systems during sustained periods of operation up to and beyond their expected licensed lifetime. This requires that new capabilities to achieve process control be developed and eventually implemented in existing nuclear assets. It also requires that approaches be developed and proven to achieve sustainability of I&C systems throughout the period of extended operation. The strategic objective of the LWRS Program Advanced Instrumentation, Information, and Control Systems Technology R&D pathway is to establish a technical basis for new technologies needed to achieve safety and reliability of operating nuclear assets and to implement new technologies in nuclear energy systems. This will be achieved by carrying out a program of R&D to develop scientific knowledge in the areas of: • Sensors, diagnostics, and prognostics to support characterization and prediction of the effects of aging and degradation phenomena effects on critical systems, structures, and components (SSCs) • Online monitoring of SSCs and active components, generation of information, and methods to analyze and employ online monitoring information • New methods for visualization, integration, and information use to enhance state awareness and leverage expertise to achieve safer, more readily available electricity generation

Present status and subjects of research on heat removal in high conversion light water reactors

International Nuclear Information System (INIS)

Murao, Yoshio

1990-01-01

Merits of high conversion LWRs: (1) The utilization of nuclear fuel several times as much as that in LWRs is possible. The rate of effective utilization of uranium is 4-6%. (2) The active storage of plutonium is feasible. (3) The bridging to the nuclear fuel cycle industries in fast reactor age can be done. (4) These contribute to the control of plutonium storage as the partner of FBRs in fast reactor age. (5) These contribute to the flexibility of medium and long term energy strategy. The reduction of natural uranium demand by the introduction of high conversion LWRs: Assuming the scale of nuclear power facilities in 2030 as 107 million kW, and that HCLWRs are introduced from 2000, the reduction till 2100 is 13%. The features of high conversion LWRs, the effect of improving the conversion ratio by spectral hardening and so on are explained. The specification of high conversion LWRs is shown in comparison with other reactor types. The aim is the high conversion PWRs in which the same safety as conventional LWRs is ensured, and energy resources and economical efficiency are attractive. The schedule of the research and the subjects of the thermo-hydraulic engineering research are shown. (K.I.)

Plutonium recycle in LWRs - The programme of the Federal Republic of Germany

International Nuclear Information System (INIS)

1979-01-01

This paper explains that in accordance with the Federal Republic of Germany's nuclear energy policy, spent fuel from their large programme of LWRs will be stored for some years in an interim storage pool until capacity for reprocessing, which is essential on environmental grounds, becomes available after 1990. The residual fissile content (uranium and plutonium) of the spent fuel can then be recovered and either be recycled into thermal LWRs or later on be burnt in Fast Reactors. A comprehensive R, D and D programme for plutonium recycle was launched in 1971. Up to the end of 1978 about 10,000 MOX fuel rods were designed, specified, fabricated and irradiated in several nuclear power plants. The methods used in this work and the results are described

Environmentally assisted cracking in Light Water Reactors: Semiannual report, October 1994--March 1995. Volume 20

International Nuclear Information System (INIS)

Chung, H.M.; Chopra, O.K.; Gavenda, D.J.; Hins, A.G.; Kassner, T.F.; Ruther, W.E.; Shack, W.J.; Soppet, W.K.

1996-01-01

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRS) from October 1994 to March 1995. Topics that have been investigated include (a) fatigue of carbon and low-alloy steel used in reactor piping and pressure vessels, (b) EAC of Alloy 600 and 690, and (c) irradiation-assisted stress corrosion cracking (IASCC) of Type 304 SS. Fatigue tests were conducted on ferritic steels in water with several dissolvedoxygen (DO) concentrations to determine whether a slow strain rate applied during different portions of a tensile-loading cycle are equally effective in decreasing fatigue life. Tensile properties and microstructures of several heats of Alloy 600 and 690 were characterized for correlation with EAC of the alloys in simulated LWR environments. Effects of DO and electrochemical potential on susceptibility to intergranular cracking of high- and commercial-purity Type 304 SS specimens from control-blade absorber tubes and a control-blade sheath irradiated in boiling water reactors were determined in slow-strain-rate-tensile tests at 289 degrees C. Microchemical changes in the specimens were studied by Auger electron spectroscopy and scanning electron microscopy to determine whether trace impurity elements may contribute to IASCC of these materials

Summary of Research on Light Water Reactor Improvement Concepts

International Nuclear Information System (INIS)

Mowery, Alfred L.

2002-01-01

The Arms Control and Disarmament Agency of the U.S. Department of State instituted a study aimed at improving the light water reactor (LWR) fuel consumption efficiency as an alternative to fuel recycle in the late 1970s. Comparison of the neutron balance tables of an LWR (1982 design) and an 'advanced' Canada deuterium uranium (CANDU) reactor explained that the relatively low fuel efficiency of the LWR was not primarily a consequence of water moderator absorptions. Rather, the comparatively low LWR fuel efficiency resulted from its use of poison to hold down startup reactivity together with other neutron losses. The research showed that each neutron saved could reduce fuel consumption by about 5%. In a typical LWR some 5 neutrons (out of 100) were absorbed in control poisons over a cycle. There are even more parasitic and leakage neutron absorptions. The objective of the research was to find ways to minimize control, parasitic, and other neutron losses aimed at improved LWR fuel consumption. Further research developed the concept of 'putting neutrons in the bank' in 238 U early in life and 'drawing them out of the bank' late in life by burning the 239 Pu produced. Conceptual designs were explored that could both control the reactor and substantially improve fuel efficiency and minimize separative work requirements.The U.S. Department of Energy augmented its high burnup fuel program based on the research in the late 1970s. As a result of the success of this program, fuel burnup in U.S. LWRs has almost doubled in the intervening two decades

Development of methodology for the analysis of fuel behavior in light water reactor in design basis accidents

International Nuclear Information System (INIS)

Salatov, A. A.; Goncharov, A. A.; Eremenko, A. S.; Kuznetsov, V. I.; Bolnov, V. A.; Gusev, A. S.; Dolgov, A. B.; Ugryumov, A. V.

2013-01-01

The report attempts to analyze the current experience of the safety fuel for light-water reactors (LWRs) under design-basis accident conditions in terms of its compliance with international requirements for licensing nuclear power plants. The components of fuel behavior analysis methodology in design basis accidents in LWRs were considered, such as classification of design basis accidents, phenomenology of fuel behavior in design basis accidents, system of fuel safety criteria and their experimental support, applicability of used computer codes and input data for computational analysis of the fuel behavior in accidents, way of accounting for the uncertainty of calculation models and the input data. A brief history of the development of probabilistic safety analysis methodology for nuclear power plants abroad is considered. The examples of a conservative approach to safety analysis of VVER fuel and probabilistic approach to safety analysis of fuel TVS-K are performed. Actual problems in development of the methodology of analyzing the behavior of VVER fuel at the design basis accident conditions consist, according to the authors opinion, in following: 1) Development of a common methodology for analyzing the behavior of VVER fuel in the design basis accidents, implementing a realistic approach to the analysis of uncertainty - in the future it is necessary for the licensing of operating VVER fuel abroad; 2) Experimental and analytical support to the methodology: experimental studies to identify and study the characteristics of the key uncertainties of computational models of fuel and the cladding, development of computational models of key events in codes, validation code on the basis of integral experiments

Databook of the isotopic composition of spent fuel in light water reactors

International Nuclear Information System (INIS)

Naito, Yoshitaka; Kurosawa, Masayoshi; Kaneko, Toshiyuki.

1993-03-01

In the framework of the activity of the nuclide production evaluation WG in the sigma committee, we summarized the measurement data of the isotopic composition of LWR spent fuels necessary to evaluate the accuracy of the burnup calculation codes. The collected data were arranged to be classified into the irradiation history of the fuel samples, the composition of the fuel assemblies, the sampling position and the isotopic composition of the fuel samples, in order to supply the information necessary to the benchmark calculation. This report describes the data collected from the 13 LWRs including the 9 LWRs (5 PWR and 4 BWR) in Europe and the USA, the 4 LWRs (2 PWR and 2 BWR) in Japan. Finally, the study on the burnup characteristics of the U, Pu isotopes is described. (author)

Application of Fully Ceramic Microencapsulated Fuels in Light Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Gentry, Cole A [ORNL; George, Nathan M [ORNL; Maldonado, G Ivan [ORNL; Godfrey, Andrew T [ORNL; Terrani, Kurt A [ORNL; Gehin, Jess C [ORNL

2012-01-01

This study aims to perform a preliminary evaluation of the feasibility of incorporation of Fully Ceramic Microencapsulated (FCM) fuels in Light Water Reactors (LWRs). In particular pin cell, lattice, and full core analyses are carried out on FCM fuel in a pressurized water reactor. Using uranium-based fuel and transuranic (TRU) based fuel in TRistructural ISOtropic (TRISO) particle form, each fuel design was examined using the SCALE 6.1 analytical suite. In regards to the uranium-based fuel, pin cell calculations were used to determine which fuel material performed best when implemented in the fuel kernel as well as the size of the kernel and surrounding particle layers. The higher physical density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design would need roughly 12% additional fissile material in comparison to that of a standard UO2 rod in order to match the lifetime of an 18-month PWR cycle. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a color-set array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In regards to the TRU based fuel, lattice calculations were performed to determine an optimal lattice design based on reactivity behavior, pin power peaking, and isotopic content. After obtaining a satisfactory lattice design, feasibility of core designs fully loaded with TRU FCM lattices was demonstrated using the NESTLE three-dimensional core simulator.

Status of advanced technology and design for water cooled reactors: Heavy water reactors

International Nuclear Information System (INIS)

1989-07-01

In 1987 the IAEA established the International Working Group on Advanced Technologies for Water-Cooled Reactors (IWGATWR). Within the framework of the IWGATWR the IAEA Technical Report on Status of Advanced Technology and Design for Water Cooled Reactors, Part I: Light Water Reactors and Part II: Heavy Water Reactors, has been undertaken to document the major current activities and trends of technological improvement and development for future water reactors. Part I of the report dealing with Light Water Reactors (LWRs) was published in 1988 (IAEA-TECDOC-479). Part II of the report covers Heavy Water Reactors (HWRs) and has now been prepared. This report is based largely upon submissions from Member States. It has been supplemented by material from the presentations at the IAEA Technical Committee and Workshop on Progress in Heavy Water Reactor Design and Technology held in Montreal, Canada, December 6-9, 1988. It is hoped that this part of the report, containing the status of advanced heavy water reactor technology up to 1988 and ongoing development programmes will aid in disseminating information to Member States and in stimulating international cooperation. Refs, figs and tabs

EPRI program in water reactor safety

International Nuclear Information System (INIS)

Loewenstein, W.B.; Gelhaus, F.; Gopalakrishnan, A.

1975-01-01

The basis for EPRI's water reactor safety program is twofold. First is compilation and development of fundamental background data necessary for quantified light-water reactor (LWR) safety assurance appraisals. Second is development of realistic and experimentally bench-marked analytical procedures. The results are expected either to confirm the safety margins in current operating parameters, and to identify overly conservative controls, or, in some cases, to provide a basis for improvements to further minimize uncertainties in expected performance. Achievement of these objectives requires the synthesis of related current and projected experimental-analytical projects toward establishment of an experimentally-based analysis for the assurance of safety for LWRs

Light water reactor safety

CERN Document Server

Pershagen, B

2013-01-01

This book describes the principles and practices of reactor safety as applied to the design, regulation and operation of light water reactors, combining a historical approach with an up-to-date account of the safety, technology and operating experience of both pressurized water reactors and boiling water reactors. The introductory chapters set out the basic facts upon which the safety of light water reactors depend. The central section is devoted to the methods and results of safety analysis. The accidents at Three Mile Island and Chernobyl are reviewed and their implications for light wate

Material Issues of Blanket Systems for Fusion Reactors - Compatibility with Cooling Water -

Science.gov (United States)

Miwa, Yukio; Tsukada, Takashi; Jitsukawa, Shiro

Environmental assisted cracking (EAC) is one of the material issues for the reactor core components of light water power reactors(LWRs). Much experience and knowledge have been obtained about the EAC in the LWR field. They will be useful to prevent the EAC of water-cooled blanket systems of fusion reactors. For the austenitic stainless steels and the reduced-activation ferritic/martensitic steels, they clarifies that the EAC in a water-cooled blanket does not seem to be acritical issue. However, some uncertainties about influences on water temperatures, water chemistries and stress conditions may affect on the EAC. Considerations and further investigations elucidating the uncertainties are discussed.

DOE-NE Light Water Reactor Sustainability Program and EPRI Long-Term Operations Program. Joint Research and Development Plan

International Nuclear Information System (INIS)

Williams, Don

2014-01-01

Nuclear power has contributed almost 20% of the total amount of electricity generated in the United States over the past two decades. High capacity factors and low operating costs make nuclear power plants (NPPs) some of the most economical power generators available. Further, nuclear power remains the single largest contributor (nearly 70%) of non-greenhouse gas-emitting electric power generation in the United States. Even when major refurbishments are performed to extend operating life, these plants continue to represent cost-effective, low-carbon assets to the nation's electrical generation capability. By the end of 2014, about one-third of the existing domestic fleet will have passed their 40th anniversary of power operations, and about one-half of the fleet will reach the same 40-year mark within this decade. Recognizing the challenges associated with pursuing extended service life of commercial nuclear power plants, the U.S. Department of Energy's (DOE) Office of Nuclear Energy [NE] and the Electric Power Research Institute (EPRI) have established separate but complementary research and development programs (DOE-NE's Light Water Reactor Sustainability [LWRS] Program and EPRI's Long-Term Operations [LTO] Program) to address these challenges. To ensure that a proper linkage is maintained between the programs, DOE-NE and EPRI executed a memorandum of understanding in late 2010 to @@@establish guiding principles under which research activities (between LWRS and LTO) could be coordinated to the benefit of both parties.@@@ This document represents the third annual revision to the initial version (March 2011) of the plan as called for in the memorandum of understanding.

DOE-NE Light Water Reactor Sustainability Program and EPRI Long-Term Operations Program. Joint Research and Development Plan

Energy Technology Data Exchange (ETDEWEB)

Williams, Don

2014-04-01

Nuclear power has contributed almost 20% of the total amount of electricity generated in the United States over the past two decades. High capacity factors and low operating costs make nuclear power plants (NPPs) some of the most economical power generators available. Further, nuclear power remains the single largest contributor (nearly 70%) of non-greenhouse gas-emitting electric power generation in the United States. Even when major refurbishments are performed to extend operating life, these plants continue to represent cost-effective, low-carbon assets to the nation’s electrical generation capability. By the end of 2014, about one-third of the existing domestic fleet will have passed their 40th anniversary of power operations, and about one-half of the fleet will reach the same 40-year mark within this decade. Recognizing the challenges associated with pursuing extended service life of commercial nuclear power plants, the U.S. Department of Energy’s (DOE) Office of Nuclear Energy (NE) and the Electric Power Research Institute (EPRI) have established separate but complementary research and development programs (DOE-NE’s Light Water Reactor Sustainability [LWRS] Program and EPRI’s Long-Term Operations [LTO] Program) to address these challenges. To ensure that a proper linkage is maintained between the programs, DOE-NE and EPRI executed a memorandum of understanding in late 2010 to “establish guiding principles under which research activities (between LWRS and LTO) could be coordinated to the benefit of both parties.” This document represents the third annual revision to the initial version (March 2011) of the plan as called for in the memorandum of understanding.

Nuclear fuel saving assessment of poison-free control in LWRs [light water reactors

International Nuclear Information System (INIS)

Abu-Zaied, G.

1988-01-01

If neutron losses to control absorbers are to be eliminated, an alternative reactivity control system has to be introduced. Due to improved neutron economy, the fuel utilization of these other alternatives is better than for a conventional poison-controlled PWR [pressurized water reactor]. It is the objective in this work to assess the uranium savings attributable to reactivity control without poison. An investigation into the savings due to the elimination of PWR control by neutron capture has been carried out. The most important finding was that up to a 30% savings in natural uranium can be achieved if fuel to moderator ratio, V f /V m , of SSC [spectral-shift-control] core at EOC [end of cycle] is similar to the standard core V f /V m

Application of fully ceramic microencapsulated fuels in light water reactors

Energy Technology Data Exchange (ETDEWEB)

Gentry, C.; George, N.; Maldonado, I. [Dept. of Nuclear Engineering, Univ. of Tennessee-Knoxville, Knoxville, TN 37996-2300 (United States); Godfrey, A.; Terrani, K.; Gehin, J. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

2012-07-01

This study performs a preliminary evaluation of the feasibility of incorporation of Fully Ceramic Microencapsulated (FCM) fuels in light water reactors (LWRs). In particular, pin cell, lattice, and full core analyses are carried out on FCM fuel in a pressurized water reactor (PWR). Using uranium-based fuel and Pu/Np-based fuel in TRistructural isotropic (TRISO) particle form, each fuel design was examined using the SCALE 6.1 analytical suite. In regards to the uranium-based fuel, pin cell calculations were used to determine which fuel material performed best when implemented in the fuel kernel as well as the size of the kernel and surrounding particle layers. The higher fissile material density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design with 19.75% enrichment would need roughly 12% additional fissile material in comparison to that of a standard UO{sub 2} rod in order to match the lifetime of an 18-month PWR cycle. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a 'color-set' array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In regards to the Pu/Np-based fuel, lattice calculations were performed to determine an optimal lattice design based on reactivity behavior, pin power peaking, and isotopic content. After obtaining a satisfactory lattice design, the feasibility of core designs fully loaded with Pu/Np FCM lattices was demonstrated using the NESTLE three-dimensional core simulator. (authors)

Application of fully ceramic microencapsulated fuels in light water reactors

International Nuclear Information System (INIS)

Gentry, C.; George, N.; Maldonado, I.; Godfrey, A.; Terrani, K.; Gehin, J.

2012-01-01

This study performs a preliminary evaluation of the feasibility of incorporation of Fully Ceramic Microencapsulated (FCM) fuels in light water reactors (LWRs). In particular, pin cell, lattice, and full core analyses are carried out on FCM fuel in a pressurized water reactor (PWR). Using uranium-based fuel and Pu/Np-based fuel in TRistructural isotropic (TRISO) particle form, each fuel design was examined using the SCALE 6.1 analytical suite. In regards to the uranium-based fuel, pin cell calculations were used to determine which fuel material performed best when implemented in the fuel kernel as well as the size of the kernel and surrounding particle layers. The higher fissile material density of uranium mononitride (UN) proved to be favorable, while the parametric studies showed that the FCM particle fuel design with 19.75% enrichment would need roughly 12% additional fissile material in comparison to that of a standard UO 2 rod in order to match the lifetime of an 18-month PWR cycle. As part of the fuel assembly design evaluations, fresh feed lattices were modeled to analyze the within-assembly pin power peaking. Also, a 'color-set' array of assemblies was constructed to evaluate power peaking and power sharing between a once-burned and a fresh feed assembly. In regards to the Pu/Np-based fuel, lattice calculations were performed to determine an optimal lattice design based on reactivity behavior, pin power peaking, and isotopic content. After obtaining a satisfactory lattice design, the feasibility of core designs fully loaded with Pu/Np FCM lattices was demonstrated using the NESTLE three-dimensional core simulator. (authors)

Trend of fuel for light water reactors and development hereafter

International Nuclear Information System (INIS)

Ichikawa, Michio; Maru, Akira; Shimoshige, Takanori

1993-01-01

Recently, the heightening of fuel burnup has been actively advanced internationally. Its degree is different according to the policy and the economical factors in respective countries. The extension of the period of operation cycle urges high burnup in view of economy. The circumstances in USA, Europe and Japan are explained. The corrosion of zircaloy cladding is the factor of limiting fuel life. The state of corrosion in reactors is different in BWRs and PWRs, and both cases are explained. The emission of FP gas from pellets to fuel rods raises the internal pressure of the fuel rods, and affects the gap conductance between pellets and cladding tubes. In the fuel for LWRs, plutonium is formed locally and burns in pellet rim part. This rim effect is discussed. The irradiation growth of fuel rods, creep down and pellet-cladding interaction are explained. The MOX fuel for LWRs and the trend of development of new type fuel are reported. The fuel for BWRs of Hitachi Ltd. and Toshiba Corp. and Nuclear Fuel Industries Ltd., the fuel for PWRs of Mitsubishi Heavy Industries Ltd. and Nuclear fuel Industries Ltd., and the recent development of the fuel cladding tubes for LWRs are described. (K.I.)

Next generation light water reactors

International Nuclear Information System (INIS)

Omoto, Akira

1992-01-01

In the countries where the new order of nuclear reactors has ceased, the development of the light water reactors of new type has been discussed, aiming at the revival of nuclear power. Also in Japan, since it is expected that light water reactors continue to be the main power reactor for long period, the technology of light water reactors of next generation has been discussed. For the development of nuclear power, extremely long lead time is required. The light water reactors of next generation now in consideration will continue to be operated till the middle of the next century, therefore, they must take in advance sufficiently the needs of the age. The improvement of the way men and the facilities should be, the simple design, the flexibility to the trend of fuel cycle and so on are required for the light water reactors of next generation. The trend of the development of next generation light water reactors is discussed. The construction of an ABWR was started in September, 1991, as No. 6 plant in Kashiwazaki Kariwa Power Station. (K.I.)

U. S. Utility Leadership in Requirements For Passive Reactors

International Nuclear Information System (INIS)

Kim, Jcng H.; Layman, William H.

1991-01-01

Utility leadership from both U.S. utilities and international utilities, is a key element in the U. S. Advanced Light Water Reactor Program. International utilities have played a very import Design reviews by the utilities participating in the ALRR Program will ensure that all of the utility requirements are met while design work is being carried out. Our mission is to achieve NRC certification of designs that reflect the needs of the utilities and we believe that this will play an important role in the resurgence of nuclear plant construction in the United States. As stated in the Nuclear Power Oversight Committee's Strategic Plan For Building New Nuclear Power Plants : 'The extensive operating experience with today's light water reactors (LWRs), and the promise shown in recent technical developments, leads the industry to the conclusion that the next nuclear plants ordered in the United States will be advanced light water reactors (A LWRs). Two types are under development : units of large output (1300 MW) called 'evolutionary' A LWRs and units of mid-size output (600 MW) called 'Passive' A LWRs. The term 'passive' refers to the safety features which depend more on natural processes such as gravity and buoyancy than on powered equipment such as pumps

Environmentally assisted cracking in light water reactors - annual report, January-December 2001

International Nuclear Information System (INIS)

Chopra, O. K.; Chung, H. M.; Clark, R. W.; Gruber, E. E; Hiller, R. W.; Shack, W. J.; Soppet, W. K.; Strain, R. V.

2003-01-01

This report summarizes work performed by Argonne National Laboratory on fatigue and environmentally assisted cracking (EAC) in light water reactors (LWRs) from January to December 2001. Topics that have been investigated include (a) environmental effects on fatigue S-N behavior of austenitic stainless steels (SSs), (b) irradiation-assisted stress corrosion cracking (IASCC) of austenitic SSs, and (c) EAC of Alloy 600. The effects of key material and loading variables, such as strain amplitude, strain rate, temperature, dissolved oxygen (DO) level in water, and material heat treatment, on the fatigue lives of wrought and cast austenitic SSs in air and LWR environments have been evaluated. The mechanism of fatigue crack initiation in austenitic SSs in LWR environments has also been examined. The results indicate that the presence of a surface oxide film or difference in the characteristics of the oxide film has no effect on fatigue crack initiation in austenitic SSs in LWR environments. Slow-strain-rate tensile tests and post-test fractographic analyses were conducted on several model SS alloys irradiated to ∼2 x 10 21 n · cm -2 (E > 1 MeV) (∼3 dpa) in He at 289 C in the Halden reactor. The results were used to determine the influence of alloying and impurity elements on the susceptibility of these steels to IASCC. Corrosion fatigue tests were conducted on nonirradiated austenitic SSs in high-purity water at 289 C to establish the test procedure and conditions that will be used for the tests on irradiated materials. A comprehensive irradiation experiment was initiated to obtain many tensile and disk specimens irradiated under simulated pressurized water reactor conditions at ∼325 C to 5, 10, 20, and 40 dpa. Crack growth tests were completed on 30% cold-worked Alloy 600 in high-purity water under various environmental and loading conditions. The results are compared with data obtained earlier on several heats of Alloy 600 tested in high-DO water under several

Inherently safe light water reactors

International Nuclear Information System (INIS)

Ise, Takeharu

1987-01-01

Today's large nuclear power reactors of world-wise use have been designed based on the philosophy. It seems that recent less electricity demand rates, higher capital cost and the TMI accident let us acknowledge relative small and simplified nuclear plants with safer features, and that Chernobyl accident in 1983 underlines the needs of intrinsic and passive safety characteristics. In such background, several inherently safe reactor concepts have been presented abroad and domestically. First describing 'Can inherently safe reactors be designed,' then I introduce representative reactor concepts of inherently safe LWRs advocated abroad so far. All of these innovative reactors employ intrinsic and passive features in their design, as follows: (1) PIUS, an acronym for Process Inherent Ultimate Safety, or an integral PWR with passive heat sink and passive shutdown mechanism, advocated by ASEA-ATOM of Sweden. (2) MAP(Minimum Attention Plant), or a self-pressurized, natural circulation integral PWR, promoted by CE Inc. of the U.S. (3) TPS(TRIGA Power System), or a compact PWR with passive heat sink and inherent fuel characteristics of large prompt temperature coefficient, prompted by GA Technologies Inc. of the U.S. (4) PIUS-BWR, or an inherently safe BWR employing passively actuated fluid valves, in competition with PIUS, prompted by ORNL of the U.S. Then, I will describe the domestic trends in Japan and the innovative inherently safe LWRs presented domestically so far. (author)

Safeguards on MOX assemblies at LWRs

International Nuclear Information System (INIS)

Arenas Carrasco, J.; Koulikov, I.; Heinonen, O.J.; Arlt, R.; Grigoleit, K.; Clarke, R.; Swinhoe, M.

2000-01-01

Operating within the framework of the New Partnership Approach (NPA) for unirradiated MOX fuel assemblies in LWRs, the IAEA and EURATOM have gained experience in safeguarding 13 LWRs licensed to operate with MOX assemblies. In order to fulfil SIR requirements, verification methods and techniques capable of measuring MOX assemblies under water have been and are still being developed. These encompass both qualitative tests for the detection of plutonium (gross attribute tests) and quantitative tests for the measurement of the amount of plutonium (partial defect tests) and are based on gamma and neutron detection techniques. There are nine PWR and two BWR where the reactor and the spent fuel pond can be covered by the same surveillance device. These are Type I reactors where the reactor and the pond are located in the same hall. In these types of facilities relying on surveillance during the MOX refuelling is especially difficult at the BWRs due to the depth of the core pond. There are two PWR type facilities where the reactor and the spent fuel pond are located in different halls and cannot be covered by the same surveillance device (Type II). An open core camera has not been installed during refuelling and therefore indirect surveillance is currently used to survey MOX loading. Improvements are therefore required and are under consideration. After receipt at the facility, there are a few facilities which must keep the received fresh MOX fuel in wet storage, not only for a short period prior to refuelling, but for more than a year, until the next refuelling campaign. In these cases timely inspections for direct use fresh nuclear material require considerable inspection effort. Additionally, where human surveillance of core loading and finally core closure are necessary there is also a large demand for manpower. Either an agreement should be reached with the operators to delay the MOX loading until the end of the fuelling campaign, or alternative approaches should be

Enhanced cooperation between IAEA and Republic of Korea on safeguards implementation at light water reactors

International Nuclear Information System (INIS)

Park, Wan-Sou; Kim, Byung-Koo; Yim, Seuk-Soon

2001-01-01

Full text: In Korea, national inspection has been initiated from the second half of 1997. From 1999, national inspection has been carried out for all nuclear facilities in Korea. In 2000, national inspections were performed successfully in 32 nuclear facilities including 12 PWRs, 4 CANDU reactors, 10 research facilities, 4 fuel fabrication plants and others. As the national inspection system settled down, both the IAEA and Korea were looking for possible ways of cooperation for mutual benefit. It was expected that considerable saving on inspection resources as well as more effective safeguards implementation could be achieved, if more enhanced cooperation work was realized. In 1999, the IAEA and Korea agreed to establish a working group for the enhanced cooperation between both sides. A working group, composed of experts from the IAEA and ROK, reviewed several options for enhanced cooperation on LWRs in Korea and suggested a measure for implementing the current safeguards approach for LWRs with remote monitoring. The basic concepts of the Enhanced Cooperation Scheme are: 1. The SSAC shall carry out all scheduled inspections for each facility for each year, while the Agency shall carry out the annual PIV and post-PIV, and a random selection of the remaining inspections; 2, The remote monitoring (RM) data necessary for technical and safeguards review shall be shared between the Agency and SSAC; 3. The IAEA shall bear the costs of purchasing RM equipment and communication operating costs from the central hub station in Korea to Vienna; the ROK will bear the costs of installing all RM equipment and communication operating costs from each LWR to the central hub station in Korea. Typically, around 8-9 inspections are performed for one LWR per annum under current safeguards approach; 1 pre-PIV, 1 PIV, 1 post-PIV, 3-4 interim inspections, fresh fuel receipts and simultaneous inspection. RM design includes 2 digital cameras (equipment hatch and spent fuel pond), VACOSS

Effect of water chemistry improvement on flow accelerated corrosion in light-water nuclear reactor

International Nuclear Information System (INIS)

Sugino, Wataru; Ohira, Taku; Nagata, Nobuaki; Abe, Ayumi; Takiguchi, Hideki

2009-01-01

Flow Accelerated Corrosion (FAC) of Carbon Steel (CS) piping has been one of main issues in Light-Water Nuclear Reactor (LWRs). Wall thinning of CS piping due to FAC increases potential risk of pipe rupture and cost for inspection and replacement of damaged pipes. In particular, corrosion products generated by FAC of CS piping brought steam generator (SG) tube corrosion and degradation of thermal performance, when it intruded and accumulated in secondary side of PWR. To preserve SG integrity by suppressing the corrosion of CS, High-AVT chemistry (Feedwater pH9.8±0.2) has been adopted to Tsuruga-2 (1160 MWe PWR, commercial operation in 1987) in July 2005 instead of conventional Low-AVT chemistry (Feedwater pH 9.3). By the High-AVT adoption, the accumulation rate of iron in SG was reduced to one-quarter of that under conventional Low-AVT. As a result, a tendency to degradation of the SG thermal efficiency was improved. On the other hand, it was clarified that High-AVT is ineffective against Flow Accelerated Corrosion (FAC) at the region where the flow turbulence is much larger. By contrast, wall thinning of CS feed water pipes due to FAC has been successfully controlled by oxygen treatment (OT) for long time in BWRs. Because Magnetite film formed on CS surface under AVT chemistry has higher solubility and porosity in comparison with Hematite film, which is formed under OT. In this paper, behavior of the FAC under various pH and dissolved oxygen concentration are discussed based on the actual wall thinning rate of BWR and PWR plant and experimental results by FAC test-loop. And, it is clarified that the FAC is suppressed even under extremely low DO concentration such as 2ppb under AVT condition in PWR. Based on this result, we propose the oxygenated water chemistry (OWC) for PWR secondary system which can mitigate the FAC of CS piping without any adverse effect for the SG integrity. Furthermore, the applicability and effectiveness of this concept developed for FAC

Water chemistry and behavior of materials in PWRs and BWRs

Energy Technology Data Exchange (ETDEWEB)

Aaltonen, P; Hanninen, H [VTT Manufacturing Technology, Espoo (Finland)

1997-09-01

Water chemistry plays a major role in corrosion and in activity transport in NPP`s. Although a full understanding of all mechanisms involved in corrosion does not exist, controlling of the water chemistry has achieved good results in recent years. Water chemistry impacts upon the operational safety of LWR`s in two main ways: integrity of pressure boundary materials and, activity transport and out-of-core radiation fields. This paper will describe application of water chemistry control in operating reactors to prevent corrosion. Some problems experienced in LWR`s will be reviewed for the design of the nuclear heating reactors (NHR). (author). 18 refs, 10 figs, 5 tabs.

Assuring PSA technical adequacy for new advanced light water reactor designs

International Nuclear Information System (INIS)

Lutz, R.J.; Detar, H.L.; Schneider, R.E.

2012-01-01

The Probabilistic Safety Assessment (PSA) for an Advanced Light Water Reactor (ALWR) must exhibit a high level of technical adequacy, or technical quality, in order to be used as a reliable tool for making risk informed decisions concerning design and eventual operation of the plant. During the design phase, decisions on some design features may use the PSA as an input. Also, the PSA may be used as input to other operational decisions during plant design and construction including the development of procedures, development of technical specification limiting conditions for operation and scheduling of preventive maintenance activities. For the existing fleet of light water reactors (LWRs), PSA technical adequacy can be judged from wide ranging acceptance criteria such as the PRA (Probabilistic Risk Assessment) Standard in the United States of America that was developed jointly by the American Society of Mechanical Engineers (ASME) and the American Nuclear Society (ANS). However, the requirements for PRA technical adequacy in this PRA Standard assumes that the plant is built and has operation experience. Some of the requirements cannot be met for ALWRs in the design or construction phase and with no operational history. Key elements of a high level of technical adequacy include procedures, operator interviews, plant walk-downs and equipment reliability histories. The ability to include these key elements into the ALWR PSA to improve technical adequacy will progress as the ALWR progresses from the design stage through the construction stage and finally to the fuel load / pre-operational stage. As the technical adequacy becomes more robust, more confidence can be placed on risk-informed decisions that are made with the PSA. To assist in using the PSA as input to design and operational decisions in the design and construction stages of an ALWR, an addition to the ASME/ANS PRA Standard is being developed. The intent of this addition to the Standard is to provide

Advanced light-water reactors

International Nuclear Information System (INIS)

Golay, M.W.; Todreas, N.E.

1990-01-01

Environmental concerns, economics and the earth's finite store of fossil fuels argue for a resuscitation of nuclear power. The authors think improved light-water reactors incorporating passive safety features can be both safe and profitable, but only if attention is paid to economics, effective management and rigorous training methods. The experience of nearly four decades has winnowed out designs for four basic types of reactor: the heavy-water reactor (HWR), the gas-cooled rector (GCR), the liquid-metal-cooled reactor (LMR) and the light-water reactor (LWR). Each design is briefly described before the paper discusses the passive safety features of the AP-600 rector, so-called because it employs an advanced pressurized water design and generates 600 MW of power

Plutonium isotopic composition of high burnup spent fuel discharged from light water reactors

International Nuclear Information System (INIS)

Nakano, Yoshihiro; Okubo, Tsutomu

2011-01-01

Highlights: → Pu isotopic composition of fuel affects FBR core nuclear characteristics very much. → Spent fuel compositions of next generation LWRs with burnup of 70 GWd/t were obtained. → Pu isotopic composition and amount in the spent fuel with 70 GWd/t were evaluated. → Spectral shift rods of high burnup BWR increases the fissile Pu fraction of spent fuel. → Wide fuel rod pitch of high burnup PWR lowers the fissile Pu fraction of spent fuel. - Abstract: The isotopic composition and amount of plutonium (Pu) in spent fuel from a high burnup boiling water reactor (HB-BWR) and a high burnup pressurized water reactor (HB-PWR), each with an average discharge burnup of 70 GWd/t, were estimated, in order to evaluate fast breeder reactor (FBR) fuel composition in the transition period from LWRs to FBRs. The HB-BWR employs spectral shift rods and the neutron spectrum is shifted through the operation cycle. The weight fraction of fissile plutonium (Puf) isotopes to the total plutonium in HB-BWR spent fuel after 5 years cooling is 62%, which is larger than that of conventional BWRs with average burnup of 45 GWd/t, because of the spectral shift operation. The amount of Pu produced in the HB-BWR is also larger than that produced in a conventional BWR. The HB-PWR uses a wider pitch 17 x 17 fuel rod assembly to optimize neutron slowing down. The Puf fraction of HB-PWR spent fuel after 5 years cooling is 56%, which is smaller than that of conventional PWRs with average burnup of 49 GWd/t, mainly because of the wider pitch. The amount of Pu produced in the HB-PWR is also smaller than that in conventional PWRs.

Cost reduction of LWRs - The main features

International Nuclear Information System (INIS)

Board, J.A.; Norman, D.

1991-01-01

For effective generation cost reductions to be achieved while maintaining safety levels, the impact of each of the above element and their interactions must be fully assessed, together with the effects of uncertainty on commercial risk. the amount of interest during construction which must be paid can be minimised by adopting designs, output ratings (unit size), and construction methods which minimise construction time. However this must be done without unduly increasing specific capital cost. Total capital costs can be reduced by sharing design and project launching costs and front-end design and licensing costs, across a series of identical plants. The paper is reviewing costs and performance factors such as those above with the aim of identifying the strategies which might be necessary within and between countries in order to create an environment which would enable cost reductions on LWRs to be made. (author)

Light-water nuclear reactors

International Nuclear Information System (INIS)

Drevon, G.

1983-01-01

This work gives basic information on light-water reactors which is advanced enough for the reader to become familiar with the essential objectives and aspects of their design, their operation and their insertion in the industrial, economic and human environment. In view of the capital role of electric energy in the modern economy a significant place is given to electron-nuclear power stations, particularly those of the type adopted for the French programme. The work includes sixteen chapters. The first chapter relates the history and presents the various applications of light water reactors. The second refers to the general elementary knowledge of reactor physics. The third chapter deals with the high power light-water nuclear power station and thereby introduces the ensuing chapters which, up to and including chapter 13, are devoted to the components and the various aspects of the operation of power stations, in particular safety and the relationship with the environment. Chapter 14 provides information on the reactors adapted to applications other than the generation of electricity on an industrial scale. Chapter 15 shows the extent of the industrial effort devoted to light-water reactors and chapter 16 indicates the paths along which the present work is preparing the future of these reactors. The various chapters have been written to allow for separate consultation. An index of the main technical terms and a bibliography complete the work [fr

Economic analysis of self-generated plutonium recycling in light water reactor

International Nuclear Information System (INIS)

Deguchi, Morimoto; Hirabayashi, Fumio; Yumoto, Ryozo

1978-01-01

This paper describes on the economics of plutonium recycle to light water reactors (LWRs). In the situation that plutonium market does not exist, it is realistic for utilities to recycle the self-generated plutonium to their own reactors. The economic incentive to recycle self-generated plutonium, plutonium fuel fabrication penalty, and the dependence of fuel cycle cost on fuel cycle cost parameters are considered. In recycling self-generated plutonium, two alternatives for fuel element design are feasible. Those are the all-plutonium design and the island design. In the present analysis, the all-plutonium design was chosen for PWRs. The calculation of reactivity variation along with burnup for both uranium fuel and plutonium fuel was done with LASER-PNC code. Plutonium inventory and other nuclear data were calculated with CHAIN code. It is expected that equilibrium composition is reached after 5 or 6 times of recycling. For the calculation of fuel cycle cost, MITCOST code was used. The recent increase in the prices of uranium ore, enrichment and reprocessing services was taken into account. The fuel cycle cost of plutonium recycle is lower than that of uranium fuel cycle within a certain limit of plutonium fabrication penalty. It is shown that the fabrication penalty of about 1250 dollar/kgHM for each plutonium successive recycle reduces the cost difference to zero. The change in other cost components affects break-even fabrication penalty, in which the fuel cycle cost of plutonium recycle is equal to that of uranium cycle. (Kato, T.)

Wastes from selected activities in two light-water reactor fuel cycles

International Nuclear Information System (INIS)

Palmer, C.R.; Hill, O.F.

1980-07-01

This report presents projected volumes and radioactivities of wastes from the production of electrical energy using light-water reactors (LWR). The projections are based upon data developed for a recent environmental impact statement in which the transuranic wastes (i.e., those wastes containing certain long-lived alpha emitters at concentrations of at least 370 becquerels, or 10 nCi, per gram of waste) from fuel cycle activities were characterized. In addition, since the WG.7 assumed that all fuel cycle wastes except mill tailings are placed in a mined geologic repository, the nontransuranic wastes from several activities are included in the projections reported. The LWR fuel cycles considered are the LWR, once-through fuel cycle (Strategy 1), in which spent fuel is packaged in metal canisters and then isolated in geologic formations; and the LWR U/Pu recycle fuel cycle (Strategy 2), wherein spent fuel is reprocessed for recovery and recycle of uranium and plutonium in LWRs. The wastes projected for the two LWR fuel cycles are summarized. The reactor operations and decommissioning were found to dominate the rate of waste generation in each cycle. These activities account for at least 85% of the fuel cycle waste volume (not including head-end wastes) when normalized to per unit electrical energy generated. At 10 years out of reactor, however, spent fuel elements in Strategy 1 represent 98% of the fuel cycle activity but only 4% of the volume. Similarly, the packaged high-level waste, fuel hulls and hardware in Strategy 2 concentrate greater than 95% of the activity in 2% of the waste volume

Modeling of Microstructure Evolution in Austenitic Stainless Steels Irradiated Under Light Water Reactor Conditions

International Nuclear Information System (INIS)

Gan, J.; Stoller, R.E.; Was, G.S.

1998-01-01

A model for the development of microstructure during irradiation in fast reactors has been adapted for light water reactor (LWR) irradiation conditions (275 approximately 325 C, up to approximately10 dpa). The original model was based on the rate-theory, and included descriptions of the evolution of both dislocation loops and cavities. The model was modified by introducing in-cascade interstitial clustering, a term to account for the dose dependence of this clustering, and mobility of interstitial clusters. The purpose of this work was to understand microstructural development under LWR irradiation with a focus on loop nucleation and saturation of loop density. It was demonstrated that in-cascade interstitial clustering dominates loop nucleation in neutron irradiation in LWRS. Furthermore it was shown that the dose dependence of in-cascade interstitial clustering is needed to account for saturation behavior as commonly observed. Both quasi-steady-state (QSS) and non-steady-state (NSS) solutions to the rate equations were obtained. The difference between QSS and NSS treatments in the calculation of defect concentration is reduced at LWR temperature when in-cascade interstitial clustering dominates loop nucleation. The mobility of interstitial clusters was also investigated and its impact on loop density is to reduce the nucleation term. The ultimate goal of this study is to combine the evolution of microstructure and microchemistry together to account for the radiation damage in austenitic stainless steels

Programme of research and development on plutonium recycling in light-water reactors; indirect nuclear action

International Nuclear Information System (INIS)

1982-01-01

This report is a summary of the results obtained during the five-year programme 'Plutonium recycle in LWRs' (1975-79) which has been made possible after the completion of the synthesis studies on 'Control and Safety' and on the 'Environmental impact'. The work performed during the programme is reported; the contracts concluded and their main results and conclusions are summarized. The activity was devoted to studies of general interest and to R and D in order to improve the scientific and technical knowledge of plutonium recycling. General studies included an assessment of the environmental impact of plutonium recycling in the EC as well as research aimed at reducing this impact. Improvements were achieved on neutron-physics parameters of the actinides, on neutron computer codes for plutonium-fuelled LWRs, on control and safety and on the in-pile behaviour of mixed-oxide fuels. The conclusions state that, while fostering a fruitful exchange of information in the Community, the programme has contributed to demonstrating the feasibility of plutonium recycling in LWRs

Implementation of defence in depth for next generation light water reactors

International Nuclear Information System (INIS)

1997-12-01

The publication of this IAEA technical document represents the conclusion of a task, initiated in 1995, devoted to defence in depth in future reactors. It focuses mainly on the next generation of LWRs, although many general considerations may also apply to other types of reactors

A passive decay heat removal system for LWRs based on air cooling

Energy Technology Data Exchange (ETDEWEB)

Mochizuki, Hiroyasu, E-mail: mochizki@u-fukui.ac.jp [Research Institute of Nuclear Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan); Yano, Takahiro [Graduate School of Engineering, University of Fukui, 1-2-4 Kanawa-cho, Tsuruga, Fukui 914-0055 (Japan)

2015-05-15

Highlights: • A passive decay heat removal system for LWRs is discussed. • An air cooler model which condenses steam is developed. • The decay heat can be removed by air coolers with forced convection. • The dimensions of the air cooler are proposed. - Abstract: The present paper describes the capability of an air cooling system (ACS) to remove decay heat from a core of LWR such as an advanced boiling water reactor (ABWR) and a pressurized water reactor (PWR). The motivation of the present research is the Fukushima severe accident (SA) on 11 March 2011. Since emergency cooling systems using electricity were not available due to station blackout (SBO) and malfunctions, many engineers might understand that water cooling was not completely reliable. Therefore, a passive decay heat removal (DHR) system would be proposed in order to prevent such an SA under the conditions of an SBO event. The plant behaviors during the SBO are calculated using the system code NETFLOW++ for the ABWR and PWR with the ACS. Two types of air coolers (ACs) are applied for the ABWR, i.e., a steam condensing air cooler (SCAC) of which intake for heat transfer tubes is provided in the steam region, and single-phase type of which intake is provided in the water region. The DHR characteristics are calculated under the conditions of the forced air circulation and also the natural air convection. As a result of the calculations, the decay heat can be removed safely by the reasonably sized ACS when heat transfer tubes are cooled with the forced air circulation. The heat removal rate per one finned heat transfer tube is evaluated as a function of air flow rate. The heat removal rate increases as a function of the air flow rate.

Environmentally assisted cracking in light water reactors annual report January - December 2005.

Energy Technology Data Exchange (ETDEWEB)

Alexandreanu, B.; Chen, Y.; Chopra, O. K.; Chung, H. M.; Gruber, E. E.; Shack, W. J.; Soppet, W. K.

2007-08-31

This report summarizes work performed from January to December 2005 by Argonne National Laboratory on fatigue and environmentally assisted cracking in light water reactors (LWRs). Existing statistical models for estimating the fatigue life of carbon and low-alloy steels and austenitic stainless steels (SSs) as a function of material, loading, and environmental conditions were updated. Also, the ASME Code fatigue adjustment factors of 2 on stress and 20 on life were critically reviewed to assess the possible conservatism in the current choice of the margins. An approach, based on an environmental fatigue correction factor, for incorporating the effects of LWR environments into ASME Section III fatigue evaluations is discussed. The susceptibility of austenitic stainless steels and their welds to irradiation-assisted stress corrosion cracking (IASCC) is being evaluated as a function of the fluence level, water chemistry, material chemistry, and fabrication history. For this task, crack growth rate (CGR) tests and slow strain rate tensile (SSRT) tests are being conducted on various austenitic SSs irradiated in the Halden boiling water reactor. The SSRT tests are currently focused on investigating the effects of the grain boundary engineering process on the IASCC of the austenitic SSs. The CGR tests were conducted on Type 316 SSs irradiated to 0.45-3.0 dpa, and on sensitized Type 304 SS and SS weld heat-affected-zone material irradiated to 2.16 dpa. The CGR tests on materials irradiated to 2.16 dpa were followed by a fracture toughness test in a water environment. The effects of material composition, irradiation, and water chemistry on growth rates are discussed. The susceptibility of austenitic SS core internals to IASCC and void swelling is also being evaluated for pressurized water reactors. Both SSRT tests and microstructural examinations are being conducted on specimens irradiated in the BOR-60 reactor in Russia to doses up to 20 dpa. Crack growth rate data

Kiche: A simulation tool for kinetics of iodine chemistry in the containment of light water reactors under severe accident conditions (Contract research)

International Nuclear Information System (INIS)

Moriyama, Kiyofumi; Maruyama, Yu; Nakamura, Hideo

2011-03-01

An iodine chemistry simulation tool, Kiche, was developed for analyses of chemical kinetics relevant to iodine volatilization in the containment vessel of light water reactors (LWRs) during a severe accident. It consists of a Fortran code to solve chemical kinetics models, reaction databases written in plain text format, and peripheral tools to convert the reaction databases into Fortran codes to solve corresponding ordinary differential equation sets. Potential advantages of Kiche are the text format reaction database separated from the code that provides flexibility of the chemistry model, and, being a Fortran code which is relatively easily coupled with other Fortran codes such as severe accident analysis codes. This document describes the model, solution method, code structure, and examples of application of Kiche for simulation of experiments. The calculation results by the present model agreed well with the experimental data and it indicates the model properly includes the most important processes in the volatilization of iodine from irradiated iodide solutions with or without organic impurities. The appendixes give practical information for the usage of Kiche. (author)

The Integrated Approach to the Accident Evaluation for Advanced LWRs

International Nuclear Information System (INIS)

Oriolo, F.; Paci, S.

1998-01-01

The present paper discusses some relevant phenomena occurring in advanced LWRs during postulated accident scenarios. In particular, the operation of ESF is the starting point for analysis of those phenomena that cause the mutual influence between PS and containment in these plants. As a consequence, it is highlighted as accident analyses which treat PS and containment phenomena completely separated may be not adequate when applied to innovate reactors. Exemplified thermal-hydraulic analysis are presented for AP600 and SBWR, using the FUMO integrated model, for highlight accident evolution taking into account these interactions. The architecture of this integrated code is presented highlighting the importance of an integrated approach to the safety analysis in innovative reactors. (author)

Determination of heavy water in heavy water - light water mixtures

International Nuclear Information System (INIS)

Sanhueza M, A.

1986-01-01

A description about experimental methodology to determine isotopic composition of heavy water - light water mixtures is presented. The employed methods are Nuclear Magnetic Resonance Spectroscopy, for measuring heavy water concentrations from 0 to 100% with intervals of 10% approx., and mass Spectrometry, for measuring heavy water concentrations from 0.1 to 1% with intervals of 0.15% approx., by means of an indirect method of Dilution. (Author)

Standards for heavy water concentration determinations in light water

International Nuclear Information System (INIS)

Varlam, M.; Steflea, D.; Pavelescu, M.

1995-01-01

The paper presents a method to prepare heavy water -light water standards within the range 144 ppm - 1%. A formula for computing standards concentration based on initial concentration of D 2 O and distilled water is given

Development of threshold action criteria for light water reactors

International Nuclear Information System (INIS)

Okrent, D.; Baldewicz, W.L.

1982-06-01

A survey of recently threshold criteria for regulatory action on LWRs is presented together with some commentary. This is followed by a new proposal for threshold action criteria which includes some different risk attributes than are found in previous criteria. Some preliminary risk values are suggested for the criteria and then evaluated in terms of a few hypothetical accident scenarios. Finally, several licensing issues are examined in terms of various threshold action criteria

Influence of fuel assembly loading pattern and fuel burnups upon leakage neutron flux spectra from light water reactor core (Joint research)

International Nuclear Information System (INIS)

Kojima, Kensuke; Okumura, Keisuke; Kosako, Kazuaki; Torii, Kazutaka

2016-01-01

At the decommissioning of light water reactors (LWRs), it is important to evaluate an amount of radioactivity in the ex-core structures such as a reactor containment vessel, radiation shieldings, and so on. It is thought that the leakage neutron spectra in these radioactivation regions, which strongly affect the induced radioactivity, would be changed by different reactor core configurations such as fuel assembly loading pattern and fuel burnups. This study was intended to evaluate these effects. For this purpose, firstly, partial neutron currents on the core surfaces were calculated for some core configurations. Then, the leakage neutron flux spectra in major radioactivation regions were calculated based on the provided currents. Finally, influence of the core configurations upon the neutron flux spectra was evaluated. As a result, it has been found that the influence is small on the spectrum shapes of neutron fluxes. However, it is necessary to pay attention to the facts that intensities of the leakage neutron fluxes are changed by the configurations and that intensities and spectrum shapes of the leakage neutron fluxes are changed depending on the angular direction around the core. (author)

A comparison of the risk measures between VHTR and LWR

International Nuclear Information System (INIS)

Han, Seok-Jung; Yang, Joon-Eon; Lee, Won-Jea

2007-01-01

Because the safety characteristics of a very high temperature reactor (VHTR) are different to that of light water reactors (LWRs), it is necessary to develop an adequate probabilistic safety assessment (PSA) methodology in order to perform a risk assessment. The inherent safety features of the VHTR are (1) simplified safety functions (2) the absence of the large release of radioactive materials such as a severe accident in LWRs. The PSA methodology for LWRs cannot be directly applied in a VHTR PSA. This paper proposes a PSA methodology for a VHTR. The essential point of the proposed methodology is to define end states of accident sequences in order to establish the risk measures for a VHTR PSA. This paper compares them with that for LWRs to discuss the differences of them

Development of light water reactors and subjects for hereafter

International Nuclear Information System (INIS)

Murao, Yoshio

1995-01-01

As for light water reactors, the structure is relatively simple, and the power plants of large capacity can be realized easily, therefore, they have been used for long period as main nuclear reactors. During that period, the accumulation of experiences on the design, manufacture, operation, maintenance and regulation of light water has become enormous, and in Japan, the social base for maintaining and developing light water reactor technologies has been prepared sufficiently. If the nuclear power generation using seawater uranium is considered, the utilization of uranium for light water reactor technologies can become the method of producing the own energy for Japan. As the factors that threaten the social base of light water reactor technologies, there are a the lowering of the desire to promote light water reactors, the effect of secular deterioration, the price rise of uranium resources, the effect of plutonium accumulation, the effect of the circumstances in developing countries and the sure recruiting of engineers. The construction and the principle of working of light water reactors and the development of light water reactors hereafter, for example, the improvement on small scale and the addition of new technology resulting in cost reduction and the lowering of the quality requirement for engineers, the improvement of core design, the countermeasures by design to serious accidents and others are described. (K.I.)

Sources of tritium

International Nuclear Information System (INIS)

Phillips, J.E.; Easterly, C.E.

1980-12-01

A review of tritium sources is presented. The tritium production and release rates are discussed for light water reactors (LWRs), heavy water reactors (HWRs), high temperature gas cooled reactors (HTGRs), liquid metal fast breeder reactors (LMFBRs), and molten salt breeder reactors (MSBRs). In addition, release rates are discussed for tritium production facilities, fuel reprocessing plants, weapons detonations, and fusion reactors. A discussion of the chemical form of the release is included. The energy producing facilities are ranked in order of increasing tritium production and release. The ranking is: HTGRs, LWRs, LMFBRs, MSBRs, and HWRs. The majority of tritium has been released in the form of tritiated water

Proceedings of the Second Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs, 23-25 September 2014, OECD-NEA HQ

International Nuclear Information System (INIS)

Massara, S.; ); Bragg-Sitton, Shannon; Pasamehmetoglu, K.; Yang, Jae Ho; Dolley, Evan J.; Rebak, Raul B.; Sowder, Andrew; Cheng, Bo; Kurata, Masaki; Van Nieuwenhove, Rudi; Li, R.; McClellan, Ken; Nelson, Andy; Carmack, Jon; Harp, Jason; Finck, Phillip; ); Kakicuhi, K.

2014-09-01

Under the guidance of the OECD-NEA Nuclear Science Committee, the expert group acts as a forum for scientific and technical information exchange on advanced light water reactor (LWR) fuels with enhanced accident tolerance. The expert group focusses on the fundamental properties and behaviour under normal operations and accident conditions for advanced core materials and components (fuels, cladding, control rods, etc.). The materials considered are applicable to Gen II and Gen III Light Water Reactors, as well as Gen III+ reactors under construction. The objective of the expert group is to define and coordinate a programme of work to help advance the scientific knowledge needed to provide the technical underpinning for the development of advanced LWR fuels with enhanced accident tolerance compared to currently used zircaloy/UO 2 fuel systems, as well as other non-fuel core components with important roles in LWR performance under accident conditions. This document brings together the available presentations (slides) given at the Second Meeting of the OECD-NEA Expert Group on Accident Tolerant Fuels for LWRs. Content: 1 - Proposed Agenda; 2 - Expert Group meeting - 23 September 2014: - Introduction and background (S. Massara, OECD-NEA) - Expected outcomes from the TFs meetings scheduled on 24-25 September (K. Pasamehmetoglu, EG Chair, INL); 3 - Task Force 1 (Systems assessment) meeting - 24 September 2014: - Metrics for the Evaluation of LWR Accident Tolerant Fuel (S. Bragg-Sitton, INL); 4 - Task Force 2 (Cladding/core materials) meeting - 24 September 2014: - Summary on SiC Task Force 2 (Clad) meeting (J.H. Yang, KAERI); - Accident Tolerant Advanced Steels Cladding for Commercial Light Water Reactors (E. Dolley, GE); - Molybdenum-Alloy Fuel Cladding Development and Testing - Update from April 2014 NEA ATF Meeting (A. Sowder, EPRI); - Accident Tolerant Control Rod Development in Japan (M. Kurata, JAEA); - IFA-774: The first in-pile test with coated fuel rods (R. Van

Error analysis of supercritical water correlations using ATHLET system code under DHT conditions

Energy Technology Data Exchange (ETDEWEB)

Samuel, J., E-mail: jeffrey.samuel@uoit.ca [Univ. of Ontario Inst. of Tech., Oshawa, ON (Canada)

2014-07-01

The thermal-hydraulic computer code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is used for analysis of anticipated and abnormal plant transients, including safety analysis of Light Water Reactors (LWRs) and Russian Graphite-Moderated High Power Channel-type Reactors (RBMKs). The range of applicability of ATHLET has been extended to supercritical water by updating the fluid-and transport-properties packages, thus enabling the code to the used in analysis of SuperCritical Water-cooled Reactors (SCWRs). Several well-known heat-transfer correlations for supercritical fluids were added to the ATHLET code and a numerical model was created to represent an experimental test section. In this work, the error in the Heat Transfer Coefficient (HTC) calculation by the ATHLET model is studied along with the ability of the various correlations to predict different heat transfer regimes. (author)

Progress in evaluation and improvement in nondestructive examination reliability for inservice inspection of Light Water Reactors (LWRs) and characterize fabrication flaws in reactor pressure vessels

International Nuclear Information System (INIS)

Doctor, S.R.; Bowey, R.E.; Good, M.S.; Friley, J.R.; Kurtz, R.J.; Simonen, F.A.; Taylor, T.T.; Heasler, P.G.; Andersen, E.S.; Diaz, A.A.; Greenwood, M.S.; Hockey, R.L.; Schuster, G.J.; Spanner, J.C.; Vo, T.V.

1991-10-01

This paper is a review of the work conducted under two programs. One (NDE Reliability Program) is a multi-year program addressing the reliability of nondestructive evaluation (NDE) for the inservice inspection (ISI) of light water reactor components. This program examines the reliability of current NDE, the effectiveness of evolving technologies, and provides assessments and recommendations to ensure that the NDE is applied at the right time, in the right place with sufficient effectiveness that defects of importance to structural integrity will be reliably detected and accurately characterized. The second program (Characterizing Fabrication Flaws in Reactor Pressure Vessels) is assembling a data base to quantify the distribution of fabrication flaws that exist in US nuclear reactor pressure vessels with respect to density, size, type, and location. These programs will be discussed as two separate sections in this report. 4 refs., 7 figs

Modeling Water Clarity and Light Quality in Oceans

Directory of Open Access Journals (Sweden)

Mohamed A. Abdelrhman

2016-11-01

Full Text Available Phytoplankton is a primary producer of organic compounds, and it forms the base of the food chain in ocean waters. The concentration of phytoplankton in the water column controls water clarity and the amount and quality of light that penetrates through it. The availability of adequate light intensity is a major factor in the health of algae and phytoplankton. There is a strong negative coupling between light intensity and phytoplankton concentration (e.g., through self-shading by the cells, which reduces available light and in return affects the growth rate of the cells. Proper modeling of this coupling is essential to understand primary productivity in the oceans. This paper provides the methodology to model light intensity in the water column, which can be included in relevant water quality models. The methodology implements relationships from bio-optical models, which use phytoplankton chlorophyll a (chl-a concentration as a surrogate for light attenuation, including absorption and scattering by other attenuators. The presented mathematical methodology estimates the reduction in light intensity due to absorption by pure seawater, chl-a pigment, non-algae particles (NAPs and colored dissolved organic matter (CDOM, as well as backscattering by pure seawater, phytoplankton particles and NAPs. The methods presented facilitate the prediction of the effects of various environmental and management scenarios (e.g., global warming, altered precipitation patterns, greenhouse gases on the wellbeing of phytoplankton communities in the oceans as temperature-driven chl-a changes take place.

NDE of stainless steel and on-line leak monitoring of LWRs. Annual report, October 1984-September 1985. Volume 2

International Nuclear Information System (INIS)

Kupperman, D.S.; Claytor, T.N.; Mathieson, T.; Prine, D.W.

1986-02-01

This progress report summarizes work performed by the Argonne National Laboratory and GARD, Inc. (Division of Chamberlain Mfg. Corp.) as subcontractor on NDE of stainless steel and on-line leak monitoring of LWRs during the 12 months from October 1984 to September 1985. 15 refs., 36 figs

Advances in light water reactor technologies

CERN Document Server

Saito, Takehiko; Ishiwatari, Yuki; Oka, Yoshiaki

2010-01-01

""Advances in Light Water Reactor Technologies"" focuses on the design and analysis of advanced nuclear power reactors. This volume provides readers with thorough descriptions of the general characteristics of various advanced light water reactors currently being developed worldwide. Safety, design, development and maintenance of these reactors is the main focus, with key technologies like full MOX core design, next-generation digital I&C systems and seismic design and evaluation described at length. This book is ideal for researchers and engineers working in nuclear power that are interested

Acoustic emission/flaw relationships for inservice monitoring of LWRs

International Nuclear Information System (INIS)

Hutton, P.H.; Kurtz, R.J.; Friesel, M.A.; Skorpik, J.R.; Dawson, J.F.

1991-10-01

The program concerning Acoustic Emission/Flaw Relationships for Inservice Monitoring of LWRs was initiated in FY76 with the objective of validating the application of acoustic emission (AE) to monitor nuclear reactor pressure-containing components during operation to detect cracking. The program has been supported by the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research. Research and development has been performed by Pacific Northwest Laboratory, operated for the Department of Energy by Battelle Memorial Institute. The program has shown the feasibility of continuous, on-line AE monitoring to detect crack growth and produced validated methods for applying the technology. Included are relationships for estimating flaw severity from AE data and field applications at Watts Bar Unit 1 Reactor, Limerick Unit 1 Reactor, and the High Flux Isotope Reactor. This report discusses the program scope and organization, the three program phases and the results obtained, standard and code activities, and instrumentation and software developed under this program

Reactor physics modelling of accident tolerant fuel for LWRs using ANSWERS codes

International Nuclear Information System (INIS)

Lindley, B.A.; Lillington, J.N.; Kotlyar, D.; Parks, G.T.; Petrovic, B.

2016-01-01

The majority of nuclear reactors operating in the world today and similarly the majority of near-term new build reactors will be LWRs. These currently accommodate traditional Zr clad UO_2/PuO_2 fuel designs which have an excellent performance record for normal operation. However, the events at Fukushima culminated in significant hydrogen production and hydrogen explosions, resulting from high temperature Zr/steam interaction following core uncovering for an extended period. These events have resulted in increased emphasis towards developing more accident tolerant fuels (ATFs)-clad systems, particularly for current and near-term build LWRs. R and D programmes are underway in the US and elsewhere to develop ATFs and the UK is engaging in these international programmes. Candidate advanced fuel materials include uranium nitride (UN) and uranium silicide (U_3Si_2). Candidate cladding materials include advanced stainless steel (FeCrAl) and silicon carbide. The UK has a long history in industrial fuel manufacture and fabrication for a wide range of reactor systems including LWRs. This is supported by a national infrastructure to perform experimental and theoretical R and D in fuel performance, fuel transient behaviour and reactor physics. In this paper, an analysis of the Integral Inherently Safe LWR design (I"2S-LWR), a reactor concept developed by an international collaboration led by the Georgia Institute of Technology, within a US DOE Nuclear Energy University Program (NEUP) Integrated Research Project (IRP) is considered. The analysis is performed using the ANSWERS reactor physics code WIMS and the EDF Energy core simulator PANTHER by researchers at the University of Cambridge. The I"2S-LWR is an advanced 2850 MWt integral PWR with inherent safety features. In order to enhance the safety features, the baseline fuel and cladding materials that were chosen for the I"2S-LWR design are U_3Si_2 and advanced stainless steel respectively. In addition, the I"2S-LWR design

Safety analysis of high temperature reactor cooled and moderated by supercritical light water

International Nuclear Information System (INIS)

Ishiwatari, Yuki; Oka, Yoshiaki; Koshizuka, Seiichi

2003-01-01

This paper describes 'Safety' of a high temperature supercritical light water cooled and moderated reactor (SCRLWR-H) with descending flow water rods. The safety system of the SCLWR-H is similar to that of a BWR. It consists of reactor scram, high pressure auxiliary feedwater system (AFS), low pressure core injection system (LPCI), safety relief valves (SRV), automatic depressurization system (ADS), and main steam isolation valves (MSIV). Ten types of transients and five types of accidents are analyzed using a plant transient analysis code SPRAT-DOWN. The sequences are determined referring to LWRs. At the 'Loss of load without turbine bypass' transient, the coolant density and the core power are increased by the over-pressurization, and at the same time the core flow rate is decreased by the closure of the turbine control valves. The peak cladding temperature increases to 727degC. The high temperature at this type of transient is one of the characteristics of the SCLWR-H. Conversely at 'feedwater-loss' events, the core power decrease to some extend by density feedback before the reactor scram. The peak cladding temperatures at the 'Partial loss of feedwater' transient and the 'Total loss of feedwater' accident are only 702degC and 833degC, respectively. The cladding temperature does not increase so much at the transients 'Loss of feedwater heating' and 'CR withdrawal' because of the operation of the plant control system. All the transients and accidents satisfy the satisfy criteria with good margins. The highest cladding temperatures of the transients and the accidents are 727degC and 833degC at the 'Loss of load without turbine bypass' and 'Total loss of feedwater', respectively. The duration of the high cladding temperature is very short at the transients. According to the parametric survey, the peak cladding temperature are sensitive to the parameters such as the pump coast-down time, delay of pump trip, AFS capacity, AFS delay, CR worth, and SRV setpoint

Light water lattices

International Nuclear Information System (INIS)

1962-01-01

The panel was attended by prominent physicists from most of the well-known laboratories in the field of light-water lattices, who exchanged the latest information on the status of work in their countries and discussed both the theoretical and the experimental aspects of the subjects. The supporting papers covered most problems, including criticality, resonance absorption, thermal utilization, spectrum calculations and the physics of plutonium bearing systems. Refs, figs and tabs

Neutron thermalization in light water

International Nuclear Information System (INIS)

Abbate, M.J.; Lolich, J.V.

1975-05-01

Investigations related to neutron thermalization in light water have been made. Neutron spectra under quasi-infinite-medium conditions have been measured by the time-of-flight technique and calculations were performed with different codes. Through the use of improved experimental techniques and the best known calculational techniques available, the known discrepancies between experimentals and theoretical values were below from 40% to 16%. The present disagreement is believed to be due the scattering model used (ENDF-GASKET, based on the modified Haywood II frequency spectra), that shows to be very satisfactory for poisoned light water cases. Moreover, previous experiments were completed and differential, integral and pulse-source experimental techniques were improved. Also a second step of a neutron and reactor calculation system was completed. (author)

Application of Cherenkov light observation to reactor measurements (1). Estimation of reactor power from Cherenkov light intensity

International Nuclear Information System (INIS)

Yamamoto, Keiichi; Takeuchi, Tomoaki; Kimura, Nobuaki; Ohtsuka, Noriaki; Tsuchiya, Kunihiko; Sano, Tadafumi; Nakajima, Ken; Homma, Ryohei; Kosuge, Fumiaki

2015-01-01

Development of the reactor measurement system was started to obtain the real-time in-core nuclear and thermal information, where the quantitative measurement of brightness of Cherenkov light was investigated. The system would be applied as a monitoring system in severe accidents and for the advanced operation management technology in existing LWRs. The calculation and the observation were performed to obtain the quantity of the Cherenkov light caused by the gamma and beta rays emitted from the fuels in the core of Kyoto University Research Reactor. The results indicate that the real-time reactor power can be estimated from the brightness of the Cherenkov light observed by a CCD camera. This method can also work for the estimation of the burn-up of spent fuels at commercial reactors. Since the observed brightness value of the Cherenkov light was influenced by the camera position, the optical observation method should be improved to achieve high accuracy observation. (author)

LIGHT WATER MODERATED NEUTRONIC REACTOR

Science.gov (United States)

Christy, R.F.; Weinberg, A.M.

1957-09-17

A uranium fuel reactor designed to utilize light water as a moderator is described. The reactor core is in a tank at the bottom of a substantially cylindrical cross-section pit, the core being supported by an apertured grid member and comprised of hexagonal tubes each containing a pluralily of fuel rods held in a geometrical arrangement between end caps of the tubes. The end caps are apertured to permit passage of the coolant water through the tubes and the fuel elements are aluminum clad to prevent corrosion. The tubes are hexagonally arranged in the center of the tank providing an amulus between the core and tank wall which is filled with water to serve as a reflector. In use, the entire pit and tank are filled with water in which is circulated during operation by coming in at the bottom of the tank, passing upwardly through the grid member and fuel tubes and carried off near the top of the pit, thereby picking up the heat generated by the fuel elements during the fission thereof. With this particular design the light water coolant can also be used as the moderator when the uranium is enriched by fissionable isotope to an abundance of U/sup 235/ between 0.78% and 2%.

Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part II: WWER conditions

Energy Technology Data Exchange (ETDEWEB)

Ernestova, M.; Zamboch, M. [Nuclear Research Institute, NRI, Rez (Czech Republic); Devrient, B.; Roth, A. [Framatome ANP GmbH, Erlangen (Germany); Ehrnsten, U. [VTT Industrial Systems, Espoo (Finland); Foehl, J.; Weissenberg, T. [Staatliche Materialpruefungsanstalt, MPA, Stuttgart (Germany); Gomez-Briceno, D.; Lapena, J. [Centro de Investigaciones Energeticas Medioambientales y Tecnologicas, CIEMAT, Madrid (Spain); Ritter, S.; Seifert, H.P. [Paul Scherrer Institute, PSI, Villigen (Switzerland)

2004-07-01

One of the ageing phenomena of pressure boundary components of light water reactors (LWRs) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of crack growth behavior of LAS in high-temperature water due to EAC under constant load (steady-state power operation), to study the effect of transient conditions (during operation or start-up/shut-down of a plant) using their impact on time-based and cycle-based crack growth rates and to a more detailed understanding of the acting mechanisms. Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurized water reactor (WWER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (WWER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarizes the most important crack growth results obtained under simulated WWER conditions. The influence of oxygen content and the effect of specimen size (C(T)25 versus C(T)50 specimens) on the crack growth rates are shown. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

Crack growth behaviour of low-alloy steels for pressure boundary components under transient light water reactor operating conditions - CASTOC, Part II: WWER conditions

International Nuclear Information System (INIS)

Ernestova, M.; Zamboch, M.; Devrient, B.; Roth, A.; Ehrnsten, U.; Foehl, J.; Weissenberg, T.; Gomez-Briceno, D.; Lapena, J.; Ritter, S.; Seifert, H.P.

2004-01-01

One of the ageing phenomena of pressure boundary components of light water reactors (LWRs) is environmentally-assisted cracking (EAC). The project CASTOC (5. Framework Programme of the EU) was launched September 2000 with six European partners and terminated August 2003. It focused in particular on the EAC behaviour of low-alloy steels (LAS) and to some extent to weld metal, heat affected zone and the influence of an austenitic cladding. The main objective was directed to the clarification of crack growth behavior of LAS in high-temperature water due to EAC under constant load (steady-state power operation), to study the effect of transient conditions (during operation or start-up/shut-down of a plant) using their impact on time-based and cycle-based crack growth rates and to a more detailed understanding of the acting mechanisms. Autoclave tests were performed with Western and Russian type reactor pressure vessel steels under simulated boiling water reactor (BWR)/normal water chemistry (NWC) and pressurized water reactor (WWER) conditions. The investigations were performed with fracture mechanics specimens of different sizes and geometries. The applied loading comprised cyclic loads, static loads and load spectra where the static load was periodically interrupted by partial unloading. With regard to water chemistry, the oxygen content (WWER) and impurities of sulphate and chlorides (BWR) were varied beyond allowable limits for continuous operation. The current paper summarizes the most important crack growth results obtained under simulated WWER conditions. The influence of oxygen content and the effect of specimen size (C(T)25 versus C(T)50 specimens) on the crack growth rates are shown. The results are discussed in the context of the current crack growth rate curves in the corresponding nuclear codes. (authors)

Critical heat flux experiments in a circular tube with heavy water and light water. (AWBA Development Program)

International Nuclear Information System (INIS)

Williams, C.L.; Beus, S.G.

1980-05-01

Experiments were performed to establish the critical heat flux (CHF) characteristics of heavy water and light water. Testing was performed with the up-flow of heavy and of light water within a 0.3744 inch inside diameter circular tube with 72.3 inches of heated length. Comparisons were made between heavy water and light water critical heat flux levels for the same local equilibrium quality at CHF, operating pressure, and nominal mass velocity. Results showed that heavy water CHF values were, on the average, 8 percent below the light water CHF values

Light energy dissipation under water stress conditions

International Nuclear Information System (INIS)

Stuhlfauth, T.; Scheuermann, R.; Fock, H.P.

1990-01-01

Using 14 CO 2 gas exchange and metabolite analyses, stomatal as well as total internal CO 2 uptake and evolution were estimated. Pulse modulated fluorescence was measured during induction and steady state of photosynthesis. Leaf water potential of Digitalis lanata EHRH. plants decreased to -2.5 megapascals after withholding irrigation. By osmotic adjustment, leaves remained turgid and fully exposed to irradiance even at severe water stress. Due to the stress-induced reduction of stomatal conductance, the stomatal CO 2 exchange was drastically reduced, whereas the total CO 2 uptake and evolution were less affected. Stomatal closure induced an increase in the reassimilation of internally evolved CO 2 . This CO 2 -recycling consumes a significant amount of light energy in the form of ATP and reducing equivalents. As a consequence, the metabolic demand for light energy is only reduced by about 40%, whereas net photosynthesis is diminished by about 70% under severe stress conditions. By CO 2 recycling, carbon flux, enzymatic substrate turnover and consumption of light energy were maintained at high levels, which enabled the plant to recover rapidly after rewatering. In stressed D. lanata plants a variable fluorescence quenching mechanism, termed coefficient of actinic light quenching, was observed. Besides water conservation, light energy dissipation is essential and involves regulated metabolic variations

Polarization Patterns of Transmitted Celestial Light under Wavy Water Surfaces

Directory of Open Access Journals (Sweden)

Guanhua Zhou

2017-03-01

Full Text Available This paper presents a model to describe the polarization patterns of celestial light, which includes sunlight and skylight, when refracted by wavy water surfaces. The polarization patterns and intensity distribution of refracted light through the wave water surface were calculated. The model was validated by underwater experimental measurements. The experimental and theoretical values agree well qualitatively. This work provides a quantitative description of the repolarization and transmittance of celestial light transmitted through wave water surfaces. The effects of wind speed and incident sources on the underwater refraction polarization patterns are discussed. Scattering skylight dominates the polarization patterns while direct solar light is the dominant source of the intensity of the underwater light field. Wind speed has an influence on disturbing the patterns under water.

Predicted Variations of Water Chemistry in the Primary Coolant Circuit of a Supercritical Water Reactor

International Nuclear Information System (INIS)

Yeh, Tsung-Kuang; Wang, Mei-Ya; Liu, Hong-Ming; Lee, Min

2012-09-01

In response to the demand over a higher efficiency for a nuclear power plant, various types of Generation IV nuclear reactors have been proposed. One of the new generation reactors adopts supercritical light water as the reactor coolant. While current in-service light water reactors (LWRs) bear an average thermal efficiency of 33%, the thermal efficiency of a supercritical water reactor (SCWR) could generally reach more than 44%. For LWRs, the coolants are oxidizing due to the presence of hydrogen peroxide and oxygen, and the degradation of structural materials has mainly resulted from stress corrosion cracking. Since oxygen is completely soluble in supercritical water, similar or even worse degradation phenomena are expected to appear in the structural and core components of an SCWR. To ensure proper designs of the structural components and suitable selections of the materials to meet the requirements of operation safety, it would be of great importance for the design engineers of an SCWR to be fully aware of the state of water chemistry in the primary coolant circuit (PCC). Since SCWRs are still in the stage of conceptual design and no practical data are available, a computer model was therefore developed for analyzing water chemistry variation and corrosion behavior of metallic materials in the PCC of a conceptual SCWR. In this study, a U.S. designed SCWR with a rated thermal power of 3575 MW and a coolant flow rate of 1843 kg/s was selected for investigating the variations in redox species concentration in the PCC. Our analyses indicated that the [H 2 ] and [H 2 O 2 ] at the core channel were higher than those at the other regions in the PCC of this SCWR. Due to the self-decomposition of H 2 O 2 , the core channel exhibited a lower [O 2 ] than the upper plenum. Because the middle water rod region was in parallel with the core channel region with relatively high dose rates, the [H 2 ] and [H 2 O 2 ] in this region were higher than those in the other regions

Light energy dissipation under water stress conditions

Energy Technology Data Exchange (ETDEWEB)

Stuhlfauth, T.; Scheuermann, R.; Fock, H.P. (Universitaet Kaiserslautern (West Germany))

1990-04-01

Using {sup 14}CO{sub 2} gas exchange and metabolite analyses, stomatal as well as total internal CO{sub 2} uptake and evolution were estimated. Pulse modulated fluorescence was measured during induction and steady state of photosynthesis. Leaf water potential of Digitalis lanata EHRH. plants decreased to {minus}2.5 megapascals after withholding irrigation. By osmotic adjustment, leaves remained turgid and fully exposed to irradiance even at severe water stress. Due to the stress-induced reduction of stomatal conductance, the stomatal CO{sub 2} exchange was drastically reduced, whereas the total CO{sub 2} uptake and evolution were less affected. Stomatal closure induced an increase in the reassimilation of internally evolved CO{sub 2}. This CO{sub 2}-recycling consumes a significant amount of light energy in the form of ATP and reducing equivalents. As a consequence, the metabolic demand for light energy is only reduced by about 40%, whereas net photosynthesis is diminished by about 70% under severe stress conditions. By CO{sub 2} recycling, carbon flux, enzymatic substrate turnover and consumption of light energy were maintained at high levels, which enabled the plant to recover rapidly after rewatering. In stressed D. lanata plants a variable fluorescence quenching mechanism, termed coefficient of actinic light quenching, was observed. Besides water conservation, light energy dissipation is essential and involves regulated metabolic variations.

Light scattering by particles in water theoretical and experimental foundations

CERN Document Server

Jonasz, Miroslaw

2007-01-01

Light scattering-based methods are used to characterize small particles suspended in water in a wide range of disciplines ranging from oceanography, through medicine, to industry. The scope and accuracy of these methods steadily increases with the progress in light scattering research. This book focuses on the theoretical and experimental foundations of the study and modeling of light scattering by particles in water and critically evaluates the key constraints of light scattering models. It begins with a brief review of the relevant theoretical fundamentals of the interaction of light with condensed matter, followed by an extended discussion of the basic optical properties of pure water and seawater and the physical principles that explain them. The book continues with a discussion of key optical features of the pure water/seawater and the most common components of natural waters. In order to clarify and put in focus some of the basic physical principles and most important features of the experimental data o...

Supercritical-pressure light water cooled reactors

CERN Document Server

Oka, Yoshiaki

2014-01-01

This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

Selected bibliography on pressure vessels for light-water-cooled power reactors (LWRs)

International Nuclear Information System (INIS)

Heddleson, F.A.

1975-01-01

Abstracts on LWR pressure vessels are arranged in the following categories: general, design, materials technology, fabrication techniques, inspection and testing, and failures. Author, keyword, and KWIC (keyword-in-content) indices are provided. (U.S.)

Developmental Light-Water Reactor Program

International Nuclear Information System (INIS)

Forsberg, C.W.

1989-12-01

This report summarizes the progress of the Developmental Light-Water Reactor (DLWR) Program at Oak Ridge National Laboratory in FY 1989. It also includes (1) a brief description of the program, (2) definition of goals, (3) earlier achievements, and (4) proposed future activities

A light-water detritiation project at Chalk River Laboratories

International Nuclear Information System (INIS)

Boniface, H.A.; Castillo, I.; Everatt, A.E.; Ryland, D.K.

2010-01-01

The NRU reactor rod bays is a large, open pool of water that receives hundreds of fuel rods annually, each carrying a small amount of residual tritiated heavy water. The tritium concentration of the rod bays water has risen over the years, to a level that is of concern to the operations staff and to the environment. The proposed long-term solution is to reduce the rod bays tritium concentration by direct detritiation of the water. The Combined Electrolytic-Catalytic Exchange (CECE) process is well suited to the light-water detritiation problem. With a tritium-protium separation factor greater than five, a CECE detritiation process can easily achieve the eight orders of magnitude separation required to split a tritiated light-water feed into an essentially tritium-free effluent stream and a tritiated heavy water product suitable for recycling through a heavy water upgrader. This paper describes a CECE light-water detritiation process specifically designed to reduce the tritium concentration in the NRU rod bays to an acceptable level. The conceptual design of a 600 Mg/a detritiation process has been developed and is now at the stage of project review and the beginning of detailed design. (author)

Design features of the Light Water Breeder Reactor (LWBR) which improve fuel utilization in light water reactors (LWBR development program)

International Nuclear Information System (INIS)

Hecker, H.C.; Freeman, L.B.

1981-08-01

This report surveys reactor core design features of the Light Water Breeder Reactor which make possible improved fuel utilization in light water reactor systems and breeding with the uranium-thorium fuel cycle. The impact of developing the uranium-thorium fuel cycle on utilization of nuclear fuel resources is discussed. The specific core design features related to improved fuel utilization and breeding which have been implemented in the Shippingport LWBR core are presented. These design features include a seed-blanket module with movable fuel for reactivity control, radial and axial reflcetor regions, low hafnium Zircaloy for fuel element cladding and structurals, and a closely spaced fuel rod lattice. Also included is a discussion of several design modifications which could further improve fuel utilization in future light water reactor systems. These include further development of movable fuel control, use of Zircaloy fuel rod support grids, and fuel element design modifications

Advanced Proliferation Resistant, Lower Cost, Uranium-Thorium Dioxide Fuels for Light Water Reactors (Progress report for work through June 2002, 12th quarterly report)

International Nuclear Information System (INIS)

Mac Donald, Philip Elsworth

2002-01-01

The overall objective of this NERI project is to evaluate the potential advantages and disadvantages of an optimized thorium-uranium dioxide (ThO2/UO2) fuel design for light water reactors (LWRs). The project is led by the Idaho National Engineering and Environmental Laboratory (INEEL), with the collaboration of three universities, the University of Florida, Massachusetts Institute of Technology (MIT), and Purdue University; Argonne National Laboratory; and all of the Pressurized Water Reactor (PWR) fuel vendors in the United States (Framatome, Siemens, and Westinghouse). In addition, a number of researchers at the Korean Atomic Energy Research Institute and Professor Kwangheon Park at Kyunghee University are active collaborators with Korean Ministry of Science and Technology funding. The project has been organized into five tasks: Task 1 consists of fuel cycle neutronics and economics analysis to determine the economic viability of various ThO2/UO2 fuel designs in PWRs; Task 2 will determine whether or not ThO2/UO2 fuel can be manufactured economically; Task 3 will evaluate the behavior of ThO2/UO2 fuel during normal, off-normal, and accident conditions and compare the results with the results of previous UO2 fuel evaluations and U.S. Nuclear Regulatory Commission (NRC) licensing standards; Task 4 will determine the long-term stability of ThO2/UO2 high-level waste; and Task 5 consists of the Korean work on core design, fuel performance analysis, and xenon diffusivity measurements

Assessment of Core Failure Limits for Light Water Reactor Fuel under Reactivity Initiated Accidents

International Nuclear Information System (INIS)

Jernkvist, Lars Olof; Massih, Ali R.

2004-12-01

significantly lower enthalpies than those required for melting, when the fuel burnup exceeds approximately 40 MWd/kgU. This issue is investigated by reviewing all high-burnup UO 2 fuel rods that have failed in RIA simulation tests in the Japanese Nuclear Safety Research Reactor and the French CABRI pulse reactor to date. Data from thirteen failed rods, with burnups between 44 and 64 MWd/kgU, indicate that clad tube failure does not necessarily lead to fuel pellet dispersal. In fact, the data suggest that a peak fuel radial average enthalpy of at least 500 J/gUO 2 is required to expel a significant part (>10 %) of the fuel inventory into the coolant. However, this empirical enthalpy threshold for fuel dispersal from high-burnup fuel rods cannot be directly applied to light water reactors, since the power pulses and/or the cooling conditions used in the pulse tests differ notably from those expected in LWRs under RIA

Fundamentals of Melt-Water Interfacial Transport Phenomena: Improved Understanding for Innovative Safety Technologies in ALWRs

Energy Technology Data Exchange (ETDEWEB)

M. Anderson; M. Corradini; K.Y. Bank; R. Bonazza; D. Cho

2005-04-26

The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The goal of this work is to provide the fundamental understanding needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability. The work considers the ex-vessel coolability phenomena in two stages. The first stage is the melt quenching process and is being addressed by Argonne National Lab and University of Wisconsin in modified test facilities. Given a quenched melt in the form of solidified debris, the second stage is to characterize the long-term debris cooling process and is being addressed by Korean Maritime University in via test and analyses. We then address the appropriate scaling and design methodologies for reactor applications.

Fundamentals of Melt-Water Interfacial Transport Phenomena: Improved Understanding for Innovative Safety Technologies in ALWRs

International Nuclear Information System (INIS)

Anderson, M.; Corradini, M.; Bank, K.Y.; Bonazza, R.; Cho, D.

2005-01-01

The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The goal of this work is to provide the fundamental understanding needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability. The work considers the ex-vessel coolability phenomena in two stages. The first stage is the melt quenching process and is being addressed by Argonne National Lab and University of Wisconsin in modified test facilities. Given a quenched melt in the form of solidified debris, the second stage is to characterize the long-term debris cooling process and is being addressed by Korean Maritime University in via test and analyses. We then address the appropriate scaling and design methodologies for reactor applications

Modeling Water Clarity and Light Quality in Oceans

Science.gov (United States)

Phytoplankton is a primary producer of organic compounds, and it forms the base of the food chain in ocean waters. The concentration of phytoplankton in the water column controls water clarity and the amount and quality of light that penetrates through it. The availability of ade...

Light-water reactor accident classification

International Nuclear Information System (INIS)

Washburn, B.W.

1980-02-01

The evolution of existing classifications and definitions of light-water reactor accidents is considered. Licensing practice and licensing trends are examined with respect to terms of art such as Class 8 and Class 9 accidents. Interim definitions, consistent with current licensing practice and the regulations, are proposed for these terms of art

Accident tolerant fuels for LWRs: A perspective

Energy Technology Data Exchange (ETDEWEB)

Zinkle, S.J., E-mail: zinklesj@ornl.gov [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States); University of Tennessee, Knoxville, TN 37996 (United States); Terrani, K.A.; Gehin, J.C.; Ott, L.J.; Snead, L.L. [Oak Ridge National Laboratory, P.O. Box 2008, Oak Ridge, TN 37831 (United States)

2014-05-01

The motivation for exploring the potential development of accident tolerant fuels in light water reactors to replace existing Zr alloy clad monolithic (U, Pu) oxide fuel is outlined. The evaluation includes a brief review of core degradation processes under design-basis and beyond-design-basis transient conditions. Three general strategies for accident tolerant fuels are being explored: modification of current state-of-the-art zirconium alloy cladding to further improve oxidation resistance (including use of coatings), replacement of Zr alloy cladding with an alternative oxidation-resistant high-performance cladding, and replacement of the monolithic ceramic oxide fuel with alternative fuel forms.

Accident tolerant fuels for LWRs: A perspective

International Nuclear Information System (INIS)

Zinkle, S.J.; Terrani, K.A.; Gehin, J.C.; Ott, L.J.; Snead, L.L.

2014-01-01

The motivation for exploring the potential development of accident tolerant fuels in light water reactors to replace existing Zr alloy clad monolithic (U, Pu) oxide fuel is outlined. The evaluation includes a brief review of core degradation processes under design-basis and beyond-design-basis transient conditions. Three general strategies for accident tolerant fuels are being explored: modification of current state-of-the-art zirconium alloy cladding to further improve oxidation resistance (including use of coatings), replacement of Zr alloy cladding with an alternative oxidation-resistant high-performance cladding, and replacement of the monolithic ceramic oxide fuel with alternative fuel forms

National need for utilizing nuclear energy for process heat generation

International Nuclear Information System (INIS)

Gambill, W.R.; Kasten, P.R.

1984-01-01

Nuclear reactors are potential sources for generating process heat, and their applications for such use economically competitive. They help satisfy national needs by helping conserve and extend oil and natural gas resources, thus reducing energy imports and easing future international energy concerns. Several reactor types can be utilized for generating nuclear process heat; those considered here are light water reactors (LWRs), heavy water reactors (HWRs), gas-cooled reactors (GCRs), and liquid metal reactors (LMRs). LWRs and HWRs can generate process heat up to 280 0 C, LMRs up to 540 0 C, and GCRs up to 950 0 C. Based on the studies considered here, the estimated process heat markets and the associated energy markets which would be supplied by the various reactor types are summarized

Fuel to Moderator Ratio Sensitivity Study Using Water Rod Moderator in SCWR Conceptual Core Design

International Nuclear Information System (INIS)

Bae, Seong Man; Kim, Yong Bae; Park, Dong Hwan; Lee, Kwang Ho

2009-01-01

The conceptual operating condition of Super Critical Water-cooled Reactor (SCWR) is above critical point of water, such that the coolant temperature ranges from 280 .deg. C to 510 .deg. C with a pressure of 25MPa. This operating condition makes an SCWR have both merits and demerits when compared with current Light Water Reactors (LWRs). One of the demerits of SCWR is degradation of neutron moderation due to a lower water density from ∼0.1g/cm 3 to ∼0.7g/cm 3 under a high coolant temperature condition. Therefore it is necessary to enhance the moderation capability for SCWR to slow down the fast fission neutrons. Many SCWR designs have a water rod concept as an additional moderator, because water has a good moderation capability. Through reviewing the previous water rod assembly designs, it was identified that a sensitivity study is required to optimize fuel assembly pitch to increase the neutron economy. In this paper, the results of the conceptual assembly design sensitivity study which focuses on the comparison of sensitivity for the fuel pitch to diameter (P/D) ratio using water rod moderator, are presented

The transuranic mass balance during the introduction of metal fuel FBR cycle

International Nuclear Information System (INIS)

Yokoo, Takeshi; Inoue, Tadashi

1999-01-01

The mass flow of plutonium and minor actinides is calculated for a future light water reactor-fast breeder reactor (LWR-FBR) transition scenario, in which power generation by LWRs is continued on a certain scale for a long period before the replacement by FBRs begins. The burnup of the LWR spent fuel is considered to be higher than the current standard. It is assumed that all the plutonium and minor actinides recovered from LWRs are used to start up and feed metal fuel commercial FBRs, which replace those LWRs that have reached the end of their life. The results show that the accumulated plutonium and minor actinides from the LWRs can be consistently consumed without further accumulation, by gradually establishing the FBR power generation and its fuel cycle on the same scale. The optimum content of the minor actinides in the standard FBR fuel is about 2 weight percents. This result indicates that if FBRs are introduced in the future, extension of the LWR usage period will cause no significant problems in terms of the consumption of accumulated transuranic elements. (author)

Recent development of nuclear power in Japan and instrumentation and control system and control room equipment for advanced light water reactors

International Nuclear Information System (INIS)

Wakayama, N.

1992-01-01

This paper was provided for the 13th IAEA/IWG-NPPCI Meeting and aims to introduce an outline of recent development of nuclear power in Japan and some topics in the field of nuclear power plant control and instrumentation. Forty units of nuclear power plants are in operation in Japan and five units of BWRs and six PWRs are under construction. Construction of prototype FBR Monju have almost completed an construction of High-Temperature Engineering Test Reactor, HTTR, started in March 1991. In parallel of those, extensive effort has been carried out to develop the third generation LWRs which are called Advanced BWR (ABWR) and Advanced PWR (APWR). Two Advanced BWRs are under safety review for construction. Instrumentation and control system of these Advanced LWRs adopts integrated digital I and C system, optical multiplexing signal transmission, fault tolerant control systems and software logic for reactor protection and safety systems and enhances plant control performance and provides human-friendly operation and maintenance environments. Main control room of these Advanced LWRs, comprised with large display panels and advanced console, has special futures such as one-man sit-down operation, human friendly man-machine interface, high level automation in operation and maintenance. (author). 7 refs, 9 figs, 1 tab

Photoelectrochemical water splitting: optimizing interfaces and light absorption

NARCIS (Netherlands)

Park, Sun-Young

2015-01-01

In this thesis several photoelectrochemical water splitting devices based on semiconductor materials were investigated. The aim was the design, characterization, and fabrication of solar-to-fuel devices which can absorb solar light and split water to produce hydrogen.

Advanced light water reactors for the nineties

International Nuclear Information System (INIS)

Ross, F.A.; Sugnet, W.R.

1987-01-01

The EPRI/Industry advanced light water reactor (ALWR) program and the US Department of Energy ALWR program are closely coordinated to meet the common objective which is the availability of improved and simplified light water reactor plants that may be ordered in the next decade to meet new or replacement capacity requirements. The EPRI/Industry objectives, program participants, and foreign participants, utility requirements document, its organization and content, small plant conceptual design program, the DOE ALWR program, design verification program, General Electric ABWR design features, Combustion Engineering system design, mid-size plant development, General Electric SBWR objectives, Westinghouse/Burns and Roe design objectives, construction improvement, and improved instrumentation and control are discussed in the paper

Systems design of direct-cycle supercritical-water-cooled fast reactors

International Nuclear Information System (INIS)

Oka, Yoshiaki; Koshizuka, Seiichi; Jevremovic, Tatjana; Okano, Yashushi

1995-01-01

The system design of a direct-cycle supercritical-water-cooled fast reactor is presented. The supercritical water does not exhibit a change of phase. the recirculation system, steam separator, and dryer of a boiling water reactor (BWR) are unnecessary. Roughly speaking, the reactor pressure vessel and control rods are similar to those of a pressurized water reactor, the containment and emergency core cooling system are similar to a BWR, and the balance of plant is similar to a supercritical-pressure fossil-fired power plant (FPP). the electric power of the fast converter is 1,508 MW(electric). The number of coolant loops is only two because of the high coolant enthalpy. Containment volume is much reduced. The thermal efficiency is improved 24% over a BWR. The coolant void reactivity is negative by placing thin zirconium-hydride layers between seeds and blankets. The power costs would be much reduced compared with those of a light water reactor (LWR) and a liquid-metal fast breeder reactor. The concept is based on the huge amount of experience with the water coolant technology of LWRs and FPPs. The oxidation of stainless steel cladding is avoided by adopting a much lower coolant temperature than that of the FPP

Materials technologies of light water reactors

International Nuclear Information System (INIS)

Begley, R.

1984-01-01

Satisfactory materials performance is a key element in achieving reliable operation of light water reactors. Outstanding performance under rigorous operational conditions has been exhibited by pressure boundary components, core internals, fuel cladding, and other critical components of these systems. Corrosion and stress corrosion phenomena have, however, had an impact on plant availability, most notably relating to pipe cracking in BWR systems and steam generator corrosion in PWR systems. These experiences have stimulated extensive development activities by the nuclear industry in improved NDE techniques, investigation of corrosion phenomena, as well as improved materials and repair processes. This paper reviews key materials performance aspects of light water reactors with particular emphasis on the progress which has been made in modeling of corrosion phenomena, control of the plant operating environment, advanced material development, and application of sophisticated repair procedures. Implementation of this technology provides the basis for improved plant availability

Thermal-Hydraulic Analysis of a Supercritical Water Reactor (SCWR) Core

International Nuclear Information System (INIS)

Kucukboyaci, V.N.; Oriani, L.

2004-01-01

The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor

Assessment Of Possible Cycle Lengths For Fully-Ceramic Micro-Encapsulated Fuel-Based Light Water Reactor Concepts

International Nuclear Information System (INIS)

Sen, R. Sonat; Pope, Michael A.; Ougouag, Abderrafi M.; Pasamehmetoglu, Kemal; Venneri, Francesco

2012-01-01

The use of TRISO-particle-based dispersion fuel within SiC matrix and cladding materials has the potential to allow the design of extremely safe LWRs with failure-proof fuel. This paper examines the feasibility of LWR-like cycle length for such a low enriched uranium fuel with the imposed constraint of strictly retaining the original geometry of the fuel pins and assemblies. The motivation for retaining the original geometry is to provide the ability to incorporate the fuel 'as-is' into existing LWRs while retaining their thermal-hydraulic characteristics. The feasibility of using this fuel is assessed by looking at cycle lengths and fuel failure rates. Other considerations (e.g., safety parameters, etc.) were not considered at this stage of the study. The study includes the examination of different TRISO kernel diameters without changing the coating layer thicknesses. The study shows that a naive use of UO 2 results in cycle lengths too short to be practical for existing LWR designs and operational demands. Increasing fissile inventory within the fuel compacts shows that acceptable cycle lengths can be achieved. In this study, starting with the recognized highest packing fraction practically achievable (44%), higher enrichment, larger fuel kernel sizes, and the use of higher density fuels have been evaluated. The models demonstrate cycle lengths comparable to those of ordinary LWRs. As expected, TRISO particles with extremely large kernels are shown to fail under all considered scenarios. In contrast, the designs that do not depart too drastically from those of the nominal NGNP HTR fuel TRISO particles are shown to perform satisfactorily and display a high rates of survival under all considered scenarios. Finally, it is recognized that relaxing the geometry constraint will result in satisfactory cycle lengths even using UO 2 -loaded TRISO particles-based fuel with enrichment at or below 20 w/o.

Neutron disadvantage factors in heavy water and light water reactors

International Nuclear Information System (INIS)

Pop-Jordanov, J.

1966-01-01

A number od heavy water and light water reactor cells are analyzed in this paper by applying analytical methods of neutron thermalization. Calculations done according to the one-group Amouyal-Benoist method are included in addition. Computer codes for ZUSE Z-23 computer were written by applying both methods. The obtained results of disadvantage factors are then compared to results obtained by one-group P 3 approximation and by multigroup K7-THERMOS code [sr

Advanced Proliferation Resistant, Lower Cost, Uranium-Thorium Dioxide Fuels for Light Water Reactors (Progress report for work through June 2002, 12th quarterly report)

Energy Technology Data Exchange (ETDEWEB)

Mac Donald, Philip Elsworth

2002-09-01

The overall objective of this NERI project is to evaluate the potential advantages and disadvantages of an optimized thorium-uranium dioxide (ThO2/UO2) fuel design for light water reactors (LWRs). The project is led by the Idaho National Engineering and Environmental Laboratory (INEEL), with the collaboration of three universities, the University of Florida, Massachusetts Institute of Technology (MIT), and Purdue University; Argonne National Laboratory; and all of the Pressurized Water Reactor (PWR) fuel vendors in the United States (Framatome, Siemens, and Westinghouse). In addition, a number of researchers at the Korean Atomic Energy Research Institute and Professor Kwangheon Park at Kyunghee University are active collaborators with Korean Ministry of Science and Technology funding. The project has been organized into five tasks: · Task 1 consists of fuel cycle neutronics and economics analysis to determine the economic viability of various ThO2/UO2 fuel designs in PWRs, · Task 2 will determine whether or not ThO2/UO2 fuel can be manufactured economically, · Task 3 will evaluate the behavior of ThO2/UO2 fuel during normal, off-normal, and accident conditions and compare the results with the results of previous UO2 fuel evaluations and U.S. Nuclear Regulatory Commission (NRC) licensing standards, · Task 4 will determine the long-term stability of ThO2/UO2 high-level waste, and · Task 5 consists of the Korean work on core design, fuel performance analysis, and xenon diffusivity measurements.

The economics of reprocessing versus direct disposal of spent nuclear fuel

International Nuclear Information System (INIS)

Bunn, M.; Holdren, J.P.; Fetter, S.; Zwaan, B. van der

2007-01-01

The economics of reprocessing versus direct disposal of spent nuclear fuel are assessed. The break-even uranium price at which reprocessing spent nuclear fuel from existing light water reactors (LWRs) and recycling the resulting plutonium and uranium in LWRs would become economic is estimated for a wide range of reprocessing prices and other fuel cycle costs and parameters. The contribution of each fuel cycle option to the cost of electricity is also estimated. A similar analysis is performed for the breakeven uranium price at which deploying fast neutron reactors (FRs) would become competitive compared with a once-through fuel cycle in LWRs, for a range of differences in capital cost between LWRs and FRs. Available information about reprocessing prices and various other fuel cycle costs and input parameters are reviewed, as well as the quantities of uranium likely to be recoverable worldwide at a range of different possible future prices. It is concluded that the once-through fuel cycle is likely to remain significantly cheaper than reprocessing and recycling in either LWRs or FRs for at least the next 50 years. Finally, there is a discussion of how scarce and expensive repository space would have to become before separation and transmutation would be economically attractive. (author)

WASH-1400: quantifying the uncertainties

International Nuclear Information System (INIS)

Erdmann, R.C.; Leverenz, F.L. Jr.; Lellouche, G.S.

1981-01-01

The purpose of this paper is to focus on the limitations of the WASH-1400 analysis in estimating the risk from light water reactors (LWRs). This assessment attempts to modify the quantification of the uncertainty in and estimate of risk as presented by the RSS (reactor safety study). 8 refs

Recent operating experience issues with 17-4 PH in LWRs

International Nuclear Information System (INIS)

Olender, A.; Gorman, J.; Marks, C.; Ilevbare, G.

2015-01-01

The stainless steel 17-4 PH has been used successfully in LWRs for quite some time. Its properties as a precipitation hardening martensitic stainless steel are desirable for high strength, high hardness applications, such as valve stems, bolting, and turbine blades. However, some in-service issues with this material have arisen over time and are addressed in this paper. Although the use of 17-4 PH in high hardness conditions is discouraged as it offers less resistance to SCC than lower hardness conditions, the utilization, and subsequent failure, of such components has occurred as recently as 2007. Thermal embrittlement can increase susceptibility to SCC, another major issue affecting 17-4 PH. The recommended maximum operating temperature is, therefore, 243 C. degrees (470 F. degrees) due to recent failures attributed to this cause. Connections with dissimilar metals have also recently been a problem, as inadequate tolerances have been left between components to allow for differences in coefficients of thermal expansion as the temperature is ramped up to operating conditions. However, with the proper precautions, 17-4 PH can, and has, performed successfully during long-term plant operation. (authors)

Light water reactor safeguards system evaluation

International Nuclear Information System (INIS)

Varnado, G.B.; Ericson, D.M. Jr.; Bennett, H.A.; Hulme, B.L.; Daniel, S.L.

1978-01-01

A methodology for assessing the effectiveness of safeguards systems was developed in this study and was applied to a typical light water reactor plant. The relative importance of detection systems, barriers, response forces and other safeguards system components was examined in extensive parameter variation studies. (author)

Light water detritiation

Energy Technology Data Exchange (ETDEWEB)

Fedorchenko, O.A.; Aleksee, I.A.; Bondarenko, S.D.; Vasyanina, T.V. [B.P. Konstantinov Petersburg Nuclear Physics Institute of National Research Centre ' Kurchatov Institute' , Gatchina (Russian Federation)

2015-03-15

Hundreds of thousands of tons of tritiated light water have been accumulating from the enterprises of nuclear fuel cycles around the world. The Dual-Temperature Water-Hydrogen (DTWH) process looks like the only practical alternative to Combined Electrolysis and Catalytic Exchange (CECE). In DTWH power-consuming lower reflux device (electrolytic cell) is replaced by a so-called 'hot tower' (LPCE column operating at conditions which ensure relatively small value of elementary separation factor α(hot)). In the upper, cold tower, the tritium transfers from hydrogen to water while in the lower, hot tower - in the opposite direction - from water to hydrogen. The DTWH process is much more complicated compared to CECE; it must be thoroughly computed and strictly controlled by an automatic control system. The use of a simulation code for DTWH is absolutely important. The simulation code EVIO-5 deals with 3 flows inside a column (hydrogen gas, water vapour and liquid water) and 2 simultaneous isotope exchange sub-processes (counter-current phase exchange and co-current catalytic exchange). EVIO-5 takes into account the strong dependence of process performance on given conditions (temperature and pressure). It calculates steady-state isotope concentration profiles considering a full set of reversible exchange reactions between different isotope modifications of water and hydrogen (12 molecular species). So the code can be used for simulation of LPCE column operation for detritiation of hydrogen and water feed, which contains H and D not only at low concentrations but above 10 at.% also. EVIO-5 code is used to model a Tritium Removal Facility with a throughput capacity of about 400 m{sup 3}/day. Simulation results show that a huge amount of wet-proofed catalyst is required (about 6000 m{sup 3}), mainly (90%) in the first stage. One reason for these large expenses (apart from a big scale of the problem itself) is the relatively high tritium separation factor in the

Light water detritiation

International Nuclear Information System (INIS)

Fedorchenko, O.A.; Aleksee, I.A.; Bondarenko, S.D.; Vasyanina, T.V.

2015-01-01

Hundreds of thousands of tons of tritiated light water have been accumulating from the enterprises of nuclear fuel cycles around the world. The Dual-Temperature Water-Hydrogen (DTWH) process looks like the only practical alternative to Combined Electrolysis and Catalytic Exchange (CECE). In DTWH power-consuming lower reflux device (electrolytic cell) is replaced by a so-called 'hot tower' (LPCE column operating at conditions which ensure relatively small value of elementary separation factor α(hot)). In the upper, cold tower, the tritium transfers from hydrogen to water while in the lower, hot tower - in the opposite direction - from water to hydrogen. The DTWH process is much more complicated compared to CECE; it must be thoroughly computed and strictly controlled by an automatic control system. The use of a simulation code for DTWH is absolutely important. The simulation code EVIO-5 deals with 3 flows inside a column (hydrogen gas, water vapour and liquid water) and 2 simultaneous isotope exchange sub-processes (counter-current phase exchange and co-current catalytic exchange). EVIO-5 takes into account the strong dependence of process performance on given conditions (temperature and pressure). It calculates steady-state isotope concentration profiles considering a full set of reversible exchange reactions between different isotope modifications of water and hydrogen (12 molecular species). So the code can be used for simulation of LPCE column operation for detritiation of hydrogen and water feed, which contains H and D not only at low concentrations but above 10 at.% also. EVIO-5 code is used to model a Tritium Removal Facility with a throughput capacity of about 400 m 3 /day. Simulation results show that a huge amount of wet-proofed catalyst is required (about 6000 m 3 ), mainly (90%) in the first stage. One reason for these large expenses (apart from a big scale of the problem itself) is the relatively high tritium separation factor in the hot tower

Neutron fluence determination for light water reactor pressure vessels

International Nuclear Information System (INIS)

Gold, R.

1994-01-01

A general description of limitations that exist in pressure vessel neutron fluence determinations for commercial light water reactors is presented. Complexity factors that arise in light water reactor pressure vessel neutron fluence calculations are identified and used to analyze calculational limitations. Two broad categories of calculational limitations are introduced, namely benchmark field limitations and deep penetration limitations. Explicit examples of limitations that can arise in each of these two broad categories are presented. These limitations are used to show that the recent draft regulatory guide for the determination of pressure vessel neutron fluence, developed by the Nuclear Regulatory Commission, is based upon procedures and assumptions that are not valid. To eliminate the complexity and limitations of calculational methods, it is recommended that the determination of light water reactor pressure vessel neutron fluence be based upon experiment. Recommendations for improved methods of pressure vessel surveillance neutron dosimetry are advanced

Flooding of a large, passive, pressure-tube light water reactor

International Nuclear Information System (INIS)

Hejzlar, P.; Todreas, N.E.; Driscoll, M.J.

1997-01-01

A reactor concept has been developed which can survive loss of coolant accidents without scram and without replenishing primary coolant inventory, while maintaining safe temperature limits on the fuel and pressure tubes. The proposed concept is a pressure tube type reactor of similar design to CANDU reactors, but differing in three key aspects. First, a solid SiC-coated graphite fuel matrix is used in place of fuel pin bundles to enable the dissipation of decay heat from the fuel in the absence of primary coolant. Second, the heavy water coolant in the pressure tubes is replaced by light water, which also serves as the moderator. Finally, the calandria tank, surrounded by a graphite reflector, contains a low pressure gas instead of heavy water moderator, and this normally-voided calandria is connected to a light water heat sink. The cover gas displaces the light water from the calandria during normal operation, while during loss of coolant or loss of heat sink accidents it allows passive calandria flooding. Calandria flooding also provides redundant and diverse reactor shutdown. This paper describes the thermal hydraulic characteristics of the passively initiated, gravity driven calandria flooding process. Flooding the calandria space with light water is a unique and very important feature of the proposed pressure-tube light water reactor (PTLWR) concept. The flooding of the top row of fuel channels must be accomplished fast enough so that in the total loss of coolant, none of the critical components of the fuel channel, i.e. the pressure tube, the calandria tube, the matrix and the fuel, exceed their design limits. The flooding process has been modeled and shown to be rapid enough to maintain all components within their design limits. (orig.)

Light and heavy water replacing system in reactor container

International Nuclear Information System (INIS)

Miyamoto, Keiji.

1979-01-01

Purpose: To enable to determine the strength of a reactor container while neglecting the outer atmospheric pressure upon evacuation, by evacuating the gap between the reactor container and a biological thermal shield, as well as the container simultaneously upon light water - heavy water replacement. Method: Upon replacing light water with heavy water by vacuum evaporation system in a nuclear reactor having a biological thermal shield surrounding the reactor container incorporating therein a reactor core by way of a heat expansion absorbing gap, the reactor container and the havy water recycling system, as well as the inside of heat expansion absorbing gap are evacuated simultaneously. This enables to neglect the outer atmospheric outer pressure upon evacuation in the determination of the container strength, and the thickness of the container can be decreased by so much as the external pressure neglected. (Moriyama, K.)

Component failures that lead to reactor scrams

International Nuclear Information System (INIS)

Burns, E.T.; Wilson, R.J.; Lim, E.Y.

1980-04-01

This report summarizes the operating experience scram data compiled from 35 operating US light water reactors (LWRs) to identify the principal components/systems related to reactor scrams. The data base utilized to identify the scram causes is developed from a EPRI-utility sponsored survey conducted by SAI coupled with recent data from the USNRC Gray Books. The reactor population considered in this evaluation is limited to 23 PWRs and 12 BWRs because of the limited scope of the program. The population includes all the US NSSS vendors. It is judged that this population accurately characterizes the component-related scrams in LWRs over the first 10 years of plant operation

Advanced fuel designs for existing and future generations of reactors: driving factors from technical and economic points of view

International Nuclear Information System (INIS)

Hesketh, Kevin

2003-01-01

This paper reviews the current state of advanced fuel research and development and considers advanced fuel development work in the context of the technical and economic drivers. The scope encompasses evolutionary development for existing light water reactors (LWRs), radical developments for LWRs, most of which are focused on more efficient plutonium consumption and on longer term developments in relation to thermal and fast reactor fuels. The review concludes that there is a gap between near-term research and development to support utilities and the long-term work that focuses on goals such as improved plutonium utilisation, waste reduction, improved proliferation resistance and strategic independence

Safety of light water reactors. Risks of nuclear technology

International Nuclear Information System (INIS)

Veser, Anke; Schlueter, Franz-Hermann; Raskob, Wolfgang; Landman, Claudia; Paesler-Sauer, Juergen; Kessler, Guenter

2012-01-01

The book on the safety of light-water reactors includes the following chapters: Part I: Physical and technical safety concept of actual German and future European light-water reactors: (1) Worldwide operated nuclear power plants in 2011, (2) Some reactor physical fundamentals. (3) Nuclear power plants in Germany. (4) Radioactive exposure due to nuclear power plants. (5) Safety concept of light-water reactors. (6) Probabilistic analyses and risk studies. (7) Design of light-water reactors against external incidents. (8) Risk comparison of nuclear power plants and other energy systems. (9) Evaluation of risk studies using the improved (new) safety concept for LWR. (19) The severe reactor accidents of Three Mile Island, Chernobyl and Fukushima. Part II: Safety of German LWR in case of a postulated aircraft impact. (11) Literature. (12) Review of requirements and actual design. (13) Incident scenarios. (14) Load approach for aircraft impact. (15) Demonstration of the structural behavior in case of aircraft impact. (16) Special considerations. (17) Evaluation of the safety state of German and foreign nuclear power plants. Part III: ROSOS as example for a computer-based decision making support system for the severe accident management. (19) Literature. (20) Radiological fundamentals, accident management, modeling of the radiological situation. (21) The decision making support system RODOS. (22) RODOS and the Fukushima accident. (23) Recent developments in the radiological emergency management in the European frame.

[Effects of light on submerged macrophytes in eutrophic water: research progress].

Science.gov (United States)

Li-Sha, Zou; Ze-Yu, Nie; Xiao-Yan, Yao; Ji-Yan, Shi

2013-07-01

The restoration of submerged macrophytes is the key to remediate eutrophic water and maintain the health of aquatic ecosystem, while light is the main limiting factor. This paper summarized the factors affecting the light extinction in water and the mechanisms of light intensity affecting the physiology of submerged macrophytes, with the focuses on the metabolic mechanisms of carbon, nitrogen, and phosphorus, the responses of antioxidant enzyme system, and the feedbacks of pigment composition and concentration in the common submerged macrophytes under low light stress. Several engineering techniques applied in the ecological restoration of submerged macrophytes were presented, and the framework of the restoration of submerged macrophytes in eutrophic water was proposed. Some problems in current research and several suggestions on future research were addressed, which could help the related research and engineering practices.

Corrosion report for the U-Mo fuel concept

Energy Technology Data Exchange (ETDEWEB)

Henager, Charles H. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Bennett, Wendy D. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Doherty, Ann L. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Fuller, E. S. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Hardy, John S. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States); Omberg, Ronald P. [Pacific Northwest National Laboratory (PNNL), Richland, WA (United States)

2014-08-28

The Fuel Cycle Research and Development (FCRD) program of the Office of Nuclear Energy (NE) has implemented a program to develop a Uranium-Molybdenum (U-Mo) metal fuel for Light Water Reactors (LWR)s. Uranium-Molybdenum fuel has the potential to provide superior performance based on its thermo-physical properties, which includes high thermal conductivity for less stored heat energy. With sufficient development, it may be able to provide the Light Water industry with a melt-resistant accident tolerant fuel with improved safety response. However, the corrosion of this fuel in reactor water environments needs to be further explored and optimized by additional alloying. The Pacific Northwest National Laboratory has been tasked with performing ex-reactor corrosion testing to characterize the performance of U-Mo fuel. This report documents the results of the effort to characterize and develop the U-Mo metal fuel concept for LWRs with regard to corrosion testing. The results of a simple screening test in buffered water at 30°C using surface alloyed U-10Mo is documented and discussed. The screening test was used to guide the selection of several potential alloy improvements that were found and are recommended for further testing in autoclaves to simulate PWR water conditions more closely.

Electrolytic separation factors for oxygen isotopes in light and heavy water solutions

International Nuclear Information System (INIS)

Gulens, J.; Olmstead, W.J.; Longhurst, T.H.; Gale, K.L.; Rolston, J.H.

1987-01-01

The electrolytic separation factor, α, has been measured for /sup 17/O and /sup 18/O at Pt and Ni anodes in both light and heavy water solutions of 6M KOH as a function of current density. For oxygen-17, isotopic separation effects were not observed, within the experimental uncertainty of +-2%, under all conditions studied. For oxygen-18, there is a small difference of 2% in α values between Pt and Ni in both light and heavy water solutions, but there is no significant difference in α values between light and heavy water solutions. In light waters solutions, the separation factor at Pt is small, α(/sup 18/O) ≤ 1.02 for i ≥ 0.1 A/cm/sub 2/. This value agrees reasonably well with theoretical estimates

The Effect of Different Perturbations on the Stability Analysis of Light Water Reactors

International Nuclear Information System (INIS)

Dykin, Victor

2010-09-01

Neutron noise analysis techniques are studied and developed, with primary use of determining the stability of Boiling Water Reactors (BWRs). In particular, the role of a specific perturbation prevailing in Light Water Reactors, the propagating density perturbation, in the stability of BWRs and on the noise field of LWRs in general, is investigated by considering three topics. In the first topics, we investigate how the neutronic response of the reactor, usually described as a second order system driven by a white noise driving force, is affected by a non-white driving force. This latter arises from the reactivity effect of the propagating density perturbations. The investigation is performed by using spectral and correlation analysis. Propagating perturbations with different velocities are analyzed. We investigate how the accuracy of the determination of the so-called decay ratio (DR) of the system, based on the assumption of white noise driving force, deteriorates with deviations from the white noise character of the driving force. In the second topics, the space dependence of the neutron noise, induced by propagating density perturbations, represented through the perturbation of the absorption, is determined and discussed. A full analytical solution was obtained by the use of the Green's function technique. The solution was analyzed for different frequencies and different system sizes. An interesting new interference effect between the point-kinetic and space-dependent components of the induced noise was discovered and interpreted in physical terms. In the last topics, a non-linear stability analysis of a BWR is performed, using so called Reduced Order Model (ROM) techniques. A ROM is usually constructed by reducing the full set of 3D space-time dependent neutron-kinetics, thermal-hydraulics and heat transfer equations to time-dependent ones, by considering space dependence in a lumped parameter model (one or two discrete channels). The main novelty of our work

Boreal Tree Light- and Water-Use: Asynchronous, Diverging, yet Complementary

Science.gov (United States)

Pappas, C.; Baltzer, J. L.; Barr, A.; Black, T. A.; Bohrer, G.; Detto, M.; Maillet, J.; Matheny, A. M.; Roy, A.; Sonnentag, O.; Stephens, J.

2017-12-01

Water stress has been suggested as a key mechanism behind the contemporary increase in tree mortality rates in northwestern North America. However, a detailed analysis of boreal tree light- and water-use strategies as well as their interspecific differences are still lacking. Here, we examine the tree hydraulic behaviour of co-occurring larch (Larix laricina) and black spruce (Picea mariana), two characteristic boreal tree species, near the southern limit of the boreal ecozone in central Canada. Sap flux density (Js) and concurrently recorded stem radius fluctuations and meteorological conditions are used to quantify tree hydraulic functioning and to scrutinize tree light- and water-use strategies. Our analysis reveals an asynchrony in the diel hydrodynamics of the two species with the initial rise in Js occurring two hours earlier in larch than in black spruce. Structural differences in the crown architecture of larch and black spruce lead to interspecific differences in light harvesting that can explain the observed asynchrony in their hydraulic function. Furthermore, the two species exhibit diverging stomatal regulation strategies with larch employing relatively isohydric whereas black spruce anisohydric behaviour. Such asynchronous and diverging tree-level light- and water-use strategies provide new insights into the ecosystem-level complementarity of tree form and function, with implications for understanding boreal forests' water and carbon dynamics and resilience to environmental stress.

Overview of light water/hydrogen-based low energy nuclear reactions

International Nuclear Information System (INIS)

Miley, George H.; Shrestha, Prajakti J.

2006-01-01

This paper reviews light water and hydrogen-based low-energy nuclear reactions (LENRs) including the different methodologies used to study these reactions and the results obtained. Reports of excess heat production, transmutation reactions, and nuclear radiation emission are cited. An aim of this review is to present a summary of the present status of light water LENR research and provide some insight into where this research is heading. (author)

Rheological Behaviour of Water-in-Light Crude Oil Emulsion

Science.gov (United States)

Husin, H.; Taju Ariffin, T. S.; Yahya, E.

2018-05-01

Basically, emulsions consist of two immiscible liquids which have different density. In petroleum industry, emulsions are undesirable due to their various costly problems in term of transportation difficulties and production loss. A study of the rheological behaviour of light crude oil and its mixture from Terengganu were carried out using Antoon Paar MCR 301 rheometer operated at pressure of 2.5 bar at temperature C. Water in oil emulsions were prepared by mixing light crude oil with different water volume fractions (20%, 30% and 40%). The objectives of present paper are to study the rheological behaviour of emulsion as a fuction of shear rate and model analysis that fitted with the experimental data. The rheological models of Ostwald-De-Waele and Herschel-Bulkley were fitted to the experimental results. All models represented well the rheological data, with high values for the correlation coefficients. The result indicated that variation of water content influenced shear rate-shear stress rheogram of the prepared emulsions. In the case of 100% light crude oil, the study demonstrated non-Newtonian shear thickening behavior. However, for emulsion with different volume water ratios, the rheological behaviour could be well described by Herschel-Bulkley models due to the present of yield stress parameter (R2 = 0.99807). As a conclusion, rheological studies showed that volume water ratio have a great impact on the shear stress and viscosity of water in oil emulsion and it is important to understand these factors to avoid various costly problems.

Utility requirements for advanced light water reactors

International Nuclear Information System (INIS)

Machiels, A.; Gray, S.; Mulford, T.; Rodwell, E.

1996-01-01

The nuclear energy industry is actively engaged in developing advanced light water reactor (ALWR) designs for the next century. The new designs take advantage of the thousands of reactor-years of experience that have been accumulated by operating over 400 plants worldwide. The EPRI effort began in the early 1980's, when a survey of utility executives was conducted to determine their prerequisites for ordering nuclear power plants. The results were clear: new plants had to be simpler and safer, and have greater design margins, i.e., be more forgiving. The utility executives also supported making improvements to the established light water reactor technology, rather than trying to develop new reactor concepts. Finally, they wanted the option to build mid-size plants (∼600 MWe) in addition to full-size plants of more than 1200 MWe. 4 refs

Light Water Reactor Generic Safety Issues Database (LWRGSIDB). User's manual

International Nuclear Information System (INIS)

1999-01-01

resolved in other plants and which can be used in reassessing the safety of individual operating plants. The IAEA-TECDOC-1044, Generic Safety Issues for Nuclear Power Plants with Light Water Reactors and Measures Taken for Their Resolution (September 1998), is a compilation of such safety issues which is based on broad international experience. This compilation is one element within the framework of IAEA activities to assist Member States in reassessing the safety of operating nuclear power plants. It is a compilation not only of the generic safety issues identified in nuclear power plants but also, in almost all cases, the measures taken to resolve these issues. The safety issues, which are generic in nature with regard to light water reactors (LWRs), and the measures taken for their resolution, are intended for use as a reference in the reassessment of the safety of operating plants.The information contained in the main body of the TECDOC has been used to establish a database. This database has search, query and report functions. This information is thus available in an electronic form which can be selectively queried and with which reports can be produced according to the requirements of the user. The database also enables the IAEA to update the data periodically on the basis of information made available by Member States

Reactor siting risk comparisons related to recommendations of NUREG-0625

International Nuclear Information System (INIS)

Barsell, A.W.; Dombek, F.S.; Orvis, D.D.

1980-11-01

This document evaluates how implementing the remote siting recommendations for nuclear reactors (NUREG-0625) made by the Siting Policy Task Force of the US Nuclear Regulatory Commission (NRC) can reduce potential public risk. The document analyzes how population density affects site-specific risk for both light water reactors (LWRs) and high-temperature gas-cooled reactors

Facilitation of decommissioning light water reactors

International Nuclear Information System (INIS)

Moore, E.B. Jr.

1979-12-01

Information on design features, special equipment, and construction methods useful in the facilitation of decommissioning light water reactors is presented. A wide range of facilitation methods - from improved documentation to special decommissioning tools and techniques - is discussed. In addition, estimates of capital costs, cost savings, and radiation dose reduction associated with these facilitation methods are given

Performance of water cooled nuclear power reactor fuels in India – Defects, failures and their mitigation

International Nuclear Information System (INIS)

Ganguly, Chaitanyamoy

2015-01-01

Water cooled and moderated nuclear power reactors account for more than 95% of the operating reactors in the world today. Light water reactors (LWRs) consisting of pressurized water reactor (PWR), their Russian counterpart namely VVER and boiling water reactor (BWR) will continue to dominate the nuclear power market. Pressurized heavy water reactor (PHWR), also known as CANDU, is the backbone of the nuclear power program in India. Updates on LWR and PHWR fuel performance are being periodically published by IAEA, OECD-NEA and the World Nuclear Association (WNA), highlighting fuel failure rate and the mitigation of fuel defects and failures. These reports clearly indicate that there has been significant improvement in in – pile fuel performance over the years and the present focus is to achieve zero fuel failure in high burn up and high performance fuels. The present paper summarizes the status of PHWR and LWR fuel performance in India, highlighting the manufacturing and the related quality control and inspection steps that are being followed at the PHWR fuel fabrication plant in order to achieve zero manufacturing defect which could contribute to achieving zero in – pile failure rate in operating and upcoming PHWR units in India. (author)

Feed water pre-heater with two steam spaces

International Nuclear Information System (INIS)

Tratz, H.; Kelp, F.; Netsch, E.

1976-01-01

A feed water pre-heater for the two stage heating of feed water by condensing steam, having a low installed height is described, which can be installed in the steam ducts of turbines of large output, as in LWRs in nuclear power stations. The inner steam space is closed on one side by the water vessel, while the tubes of the inner steam space go straight from the water vessel, and the tubes of the outer steam space are bent into a U shape and open out into the water vessel. The two-stage preheater is thus surrounded by feedwater in two ways. (UWI) [de

Current status of light water reactor and Hitachi's technical improvements for BWR

International Nuclear Information System (INIS)

Miki, Minoru; Ohki, Arahiko.

1984-01-01

Gradual technical improvements in Japan over the years has improved the reliability of light water reactors, and has achieved the highest capacity factor level in the world. Commercial operation of Fukushima 2-2 (1,100 MW) of the Tokyo Electric Power Co. was started in February, 1984, as the first standardized BWR base plant, ushering in a new age of domestic light water reactor technology. The ABWR (1,300 MW class) has been developed as Japan's next generation light water reactor, with construction aimed at the latter half of the 1980's. Hitachi's extensive efforts range from key nuclear equipment to various related robots, directed at improving safety, reliability, and the capacity factor, while reducing radiation exposure. This paper presents an outline of Hitachi's participation in the light water reactor's improvement and standardization, and the current status of our role in the international cooperation plan for the ABWR. (author)

In situ measurement of inelastic light scattering in natural waters

Science.gov (United States)

Hu, Chuanmin

Variation in the shape of solar absorption (Fraunhofer) lines are used to study the inelastic scattering in natural waters. In addition, oxygen absorption lines near 689nm are used to study the solar stimulated chlorophyll fluorescence. The prototype Oceanic Fraunhofer Line Discriminator (OFLD) has been further developed and improved by using a well protected fiber optic - wire conductor cable and underwater electronic housing. A Monte-Carlo code and a simple code have been modified to simulate the Raman scattering, DOM fluorescence and chlorophyll fluorescence. A series of in situ measurements have been conducted in clear ocean waters in the Florida Straits, in the turbid waters of Florida Bay, and in the vicinity of a coral reef in the Dry Tortugas. By comparing the reduced data with the model simulation results, the Raman scattering coefficient, b r with an excitation wavelength at 488nm, has been verified to be 2.6 × 10-4m-1 (Marshall and Smith, 1990), as opposed to 14.4 × 10- 4m-1 (Slusher and Derr, 1975). The wavelength dependence of b r cannot be accurately determined from the data set as the reported values (λ m-4 to λ m- 5) have an insignificant effect in the natural underwater light field. Generally, in clear water, the percentage of inelastic scattered light in the total light field at /lambda 510nm. At low concentrations (a y(/lambda = 380nm) less than 0.1m-1), DOM fluorescence plays a small role in the inelastic light field. However, chlorophyll fluorescence is much stronger than Raman scattering at 685nm. In shallow waters where a sea bottom affects the ambient light field, inelastic light is negligible for the whole visible band. Since Raman scattering is now well characterized, the new OFLD can be used to measure the solar stimulated in situ fluorescence. As a result, the fluorescence signals of various bottom surfaces, from coral to macrophytes, have been measured and have been found to vary with time possibly due to nonphotochemical quenching

Melt quenching and coolability by water injection from below: Co-injection of water and non-condensable gas

International Nuclear Information System (INIS)

Cho, Dae H.; Page, Richard J.; Abdulla, Sherif H.; Anderson, Mark H.; Klockow, Helge B.; Corradini, Michael L.

2006-01-01

The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific advanced light water reactor (ALWR) designs, deliberate mixing of the core melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The goal of our work is to provide the fundamental understanding needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability. The work considers the ex-vessel coolability phenomena in two stages. The first stage is the melt quenching process and is being addressed by Argonne National Lab and University of Wisconsin in modified test facilities. Given a quenched melt in the form of solidified debris, the second stage is to characterize the long-term debris cooling process and is being addressed by Korean Maritime University via test and analyses. In this paper, experiments on melt quenching by the injection of water from below are addressed. The test section represented one-dimensional flow-channel simulation of the bottom injection of water into a core melt in the reactor cavity. The melt simulant was molten lead or a lead alloy (Pb-Bi). For the experimental conditions employed (i.e., melt depth and water flow rates), it was found that: (1) the volumetric heat removal rate increased with increasing water mass flow rate and (2) the non-condensable gas mixed with the injected water had no impairing effect on the overall heat removal rate. Implications of these current experimental findings for ALWR ex-vessel coolability are discussed

Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

International Nuclear Information System (INIS)

Ragusa, Jean; Vierow, Karen

2011-01-01

The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

Analysis of Advanced Fuel Assemblies and Core Designs for the Current and Next Generations of LWRs

Energy Technology Data Exchange (ETDEWEB)

Ragusa, Jean; Vierow, Karen

2011-09-01

The objective of the project is to design and analyze advanced fuel assemblies for use in current and future light water reactors and to assess their ability to reduce the inventory of transuranic elements, while preserving operational safety. The reprocessing of spent nuclear fuel can delay or avoid the need for a second geological repository in the US. Current light water reactor fuel assembly designs under investigation could reduce the plutonium inventory of reprocessed fuel. Nevertheless, these designs are not effective in stabilizing or reducing the inventory of minor actinides. In the course of this project, we developed and analyzed advanced fuel assembly designs with improved thermal transmutation capability regarding transuranic elements and especially minor actinides. These designs will be intended for use in thermal spectrum (e.g., current and future fleet of light water reactors in the US). We investigated various fuel types, namely high burn-up advanced mixed oxides and inert matrix fuels, in various geometrical designs that are compliant with the core internals of current and future light water reactors. Neutronic/thermal hydraulic effects were included. Transmutation efficiency and safety parameters were used to rank and down-select the various designs.

Ultraviolet light: sterile water without chlorine smell and taste

International Nuclear Information System (INIS)

Anon.

1977-01-01

The use of chlorine and hypochlorite is necessary in larger waterworks, but it is a disadvantage in smaller plants, where overtreatment easily leads to smell and taste of chlorine in the water. Ultraviolet light with a wavelength of 2535 Angstrom gives 100% disinfection with a dose of 10 mWs/cm 2 for all known bacteria. In practice a dose of 40 mWs/cm 2 and an irradiation time of 15 minutes is desireable. A standard unit utilising six UV light tubes arranged concentrically around a quartz tube, through which the water flows, is described briefly. (JIW)

Ultraviolet light: sterile water without chlorine smell and taste

Energy Technology Data Exchange (ETDEWEB)

1977-02-14

The use of chlorine and hypochlorite is necessary in larger waterworks, but it is a disadvantage in smaller plants, where overtreatment easily leads to smell and taste of chlorine in the water. Ultraviolet light with a wavelength of 2535 Angstrom gives 100% disinfection with a dose of 10 mWs/cm/sup 2/ for all known bacteria. In practice a dose of 40 mWs/cm/sup 2/ and an irradiation time of 15 minutes is desireable. A standard unit utilising six UV light tubes arranged concentrically around a quartz tube, through which the water flows, is described briefly.

US Advanced Light Water Reactor Program; overall objective

International Nuclear Information System (INIS)

Klug, N.

1989-01-01

The overall objective of the US Department of Energy (DOE) Advanced Light Water Reactor (ALWR) program is to perform coordinated programs of the nuclear industry and DOE to insure the availability of licensed, improved, and simplified light water reactor standard plant designs that may be ordered in the 1990's to help meet the US electrical power demand. The discussion includes plans to meet program objectives and the design certification program. DOE is currently supporting the development of conceptual designs, configurations, arrangements, construction methods/plans, and proof test key design features for the General Electric ASBWR and the Westinghouse AP600. Key features of each are summarized. Principal milestones related to licensing of large standard plants, simplified mid-size plant development, and plant lifetime improvement are noted

OVERVIEW OF MODULAR HTGR SAFETY CHARACTERIZATION AND POSTULATED ACCIDENT BEHAVIOR LICENSING STRATEGY

Energy Technology Data Exchange (ETDEWEB)

Ball, Sydney J [ORNL

2014-06-01

This report provides an update on modular high-temperature gas-cooled reactor (HTGR) accident analyses and risk assessments. One objective of this report is to improve the characterization of the safety case to better meet current regulatory practice, which is commonly geared to address features of today s light water reactors (LWRs). The approach makes use of surrogates for accident prevention and mitigation to make comparisons with LWRs. The safety related design features of modular HTGRs are described, along with the means for rigorously characterizing accident selection and progression methodologies. Approaches commonly used in the United States and elsewhere are described, along with detailed descriptions and comments on design basis (and beyond) postulated accident sequences.

Safety and licensing for small and medium power reactors

International Nuclear Information System (INIS)

Trauger, D.B.

1987-01-01

Proposed new concepts for small and medium power reactors differ substantially from traditional Light Water Reactors (LWRs). Although designers have a large base of experience in safety and licensing, much of it is not relevant to new concepts. It can be a disadvantage if regulators apply LWR rules directly. A fresh start is appropriate. The extensive interactions between industry, regulators, and the public complicates but may enhance safety. It is basic to recognize the features that distinguish nuclear energy safety from that for other industries. These features include: nuclear reactivity, fission product radiation, and radioactive decay heat. Small and medium power reactors offer potential advantages over LWRs, particularly for reactivity and decay heat

Safety and licensing for small and medium power reactors

International Nuclear Information System (INIS)

Trauger, D.B.

1988-01-01

Proposed new concepts for small and medium power reactors differ substantially from traditional Light Water Reactors (LWRs). Although designers have a large base of experience in safety and licensing, much of it is not relevant to new concepts. It can be a disadvantage if regulators apply LWR rules directly. A fresh start is appropriate. The extensive interactions between industry, regulators, and the public complicate but may enhance safety. It is basic to recognize the features that distinguish nuclear energy safety from that for other industries. These features include: Nuclear reactivity, fission product radiation, and radioactive decay heat. Small and medium power reactors offer potential advantages over LWRs, particularly for reactivity and decay heat. (orig.)

Status report on fast reactor recycle and impact on geologic disposal

International Nuclear Information System (INIS)

Bauer, T. H.; Morris, E. E.; Wigeland, R. A.

2007-01-01

The GNEP program envisions continuing the use of light-water reactors (LWRs), with the addition of processing the discharged, or spent, LWR fuel to recover actinide and fission product elements, and then recycling the actinide elements in sodium-cooled fast reactors. Previous work has established the relationship between the processing efficiencies of spent LWR fuel, as represented by spent PWR fuel, and the potential increase in repository utilization for the resulting processing waste. The purpose of this current study is to determine a similar relationship for the waste from processing spent fast reactor fuel, and then to examine the wastes from the combination of LWRs and fast reactors as would be deployed with the GNEP approach

Exposure control during high maintenance jobs

International Nuclear Information System (INIS)

Hinson, C.S.; Wigginton, J.E.

1991-01-01

Occupational radiation doses at US light-water-cooled reactors (LWRs) have been decreasing every year since 1983. In 1989, the LWR average dose per reactor reached its lowest point since 1973. Historically, maintenance-related activities have accounted for between 70 and 75% of the total occupational doses at US LWRs, and most of this maintenance-related dose is accrued during plant outages. This paper will focus on high dose maintenance jobs such as steam generator replacement, recirculation pipe replacement/crack repair, induction heating stress improvement, and steam generator tube plugging/sleeving and will discuss how the doses associated with these jobs have declined over the past several years through the implementation of various ALARA techniques

Equations of state for light water

International Nuclear Information System (INIS)

Rubin, G.A.; Granziera, M.R.

1983-01-01

The equations of state for light water were developed, based on the tables of Keenan and Keyes. Equations are presented, describing the specific volume, internal energy, enthalpy and entropy of saturated steam, superheated vapor and subcooled liquid as a function of pressure and temperature. For each property, several equations are shown, with different precisions and different degress of complexity. (Author) [pt

Comparative evaluation of recent water hammer events in light water reactors

International Nuclear Information System (INIS)

House, R.K.; Sursock, J.P.; Kim, J.H.

1987-01-01

Water hammer events that occurred in commercial U.S. light water reactors in the five-year period from 1981 to 1985 were surveyed, and a preliminary evaluation of the events was conducted. The information developed supplements a previous study which evaluated water hammer events in the twelve-year period from 1969 to 1981. The current study of water hammer events in the 1980's confirms that the rate of events remains relatively constant (less than 0.25 events per plant year) in both PWRs and BWRs. Although water hammer events are not normally considered a safety issue, the economic impact of the events on plant operations can be significant. One particular severe water hammer event is estimated to have cost the plant owner $10 million for repair and evaluation alone. A variety of key characteristics of the recent water hammer events are summarized to establish a basis for further study of preventative methods

BWR water chemistry guidelines and PWR primary water chemistry guidelines in Japan – Purpose and technical background

Energy Technology Data Exchange (ETDEWEB)

Kawamura, Hirotaka, E-mail: kawamuh@criepi.denken.or.jp [Central Research Institute of Electric Power Industry (Japan); Hirano, Hideo [Central Research Institute of Electric Power Industry (Japan); Katsumura, Yousuke [University of Tokyo (Japan); Uchida, Shunsuke [Tohoku University (Japan); Mizuno, Takayuki [Mie University (Japan); Kitajima, Hideaki; Tsuzuki, Yasuo [Japan Nuclear Safety Institute (Japan); Terachi, Takumi [Institute of Nuclear Safety System, Inc. (Japan); Nagase, Makoto; Usui, Naoshi [Hitachi-GE Nuclear Energy, Ltd. (Japan); Takagi, Junichi; Urata, Hidehiro [Toshiba Corporation (Japan); Shoda, Yasuhiko; Nishimura, Takao [Mitsubishi Heavy Industry, Ltd. (Japan)

2016-12-01

Highlights: • Framework of BWR/PWR water chemistry Guidelines in Japan are presented. • Guideline necessity, definitions, philosophy and technical background are mentioned. • Some guideline settings for control parameters and recommendations are explaines. • Chemistry strategy is also mentioned. - Abstract: After 40 years of light water reactor (LWR) operations in Japan, the sustainable development of water chemistry technologies has aimed to ensure the highest coolant system component integrity and fuel reliability performance for maintaining LWRs in the world; additionally, it aimed to achieve an excellent dose rate reduction. Although reasonable control and diagnostic parameters are utilized by each boiling water reactor (BWR) and pressurized water reactor (PWR) owner, it is recognized that specific values are not shared among everyone involved. To ensure the reliability of BWR and PWR operation and maintenance, relevant members of the Atomic Energy Society of Japan (AESJ) decided to establish guidelines for water chemistry. The Japanese BWR and PWR water chemistry guidelines provide strategies to improve material and fuel reliability performance as well as to reduce dosing rates. The guidelines also provide reasonable “control values”, “diagnostic values” and “action levels” for multiple parameters, and they stipulate responses when these levels are exceeded. Specifically, “conditioning parameters” are adopted in the Japanese PWR primary water chemistry guidelines. Good practices for operational conditions are also discussed with reference to long-term experience. This paper presents the purpose, technical background and framework of the preliminary water chemistry guidelines for Japanese BWRs and PWRs. It is expected that the guidelines will be helpful as an introduction to achieve safety and reliability during operations.

Visible-Light-Driven BiOI-Based Janus Micromotor in Pure Water.

Science.gov (United States)

Dong, Renfeng; Hu, Yan; Wu, Yefei; Gao, Wei; Ren, Biye; Wang, Qinglong; Cai, Yuepeng

2017-02-08

Light-driven synthetic micro-/nanomotors have attracted considerable attention due to their potential applications and unique performances such as remote motion control and adjustable velocity. Utilizing harmless and renewable visible light to supply energy for micro-/nanomotors in water represents a great challenge. In view of the outstanding photocatalytic performance of bismuth oxyiodide (BiOI), visible-light-driven BiOI-based Janus micromotors have been developed, which can be activated by a broad spectrum of light, including blue and green light. Such BiOI-based Janus micromotors can be propelled by photocatalytic reactions in pure water under environmentally friendly visible light without the addition of any other chemical fuels. The remote control of photocatalytic propulsion by modulating the power of visible light is characterized by velocity and mean-square displacement analysis of optical video recordings. In addition, the self-electrophoresis mechanism has been confirmed for such visible-light-driven BiOI-based Janus micromotors by demonstrating the effects of various coated layers (e.g., Al 2 O 3 , Pt, and Au) on the velocity of motors. The successful demonstration of visible-light-driven Janus micromotors holds a great promise for future biomedical and environmental applications.

Light-water reactor research and development

International Nuclear Information System (INIS)

1985-05-01

This report on the national program of research and development on light water reactors is the second of two reports requested in 1982 by W. Kenneth Davis, Deputy Secretary of the Department of Energy. A first report, published in September 1983, treated the needs for safety-related R and D. In this second report, the Energy Research Advisory Board finds that, although many light water reactors are providing reliable and economic electricity, it appears unlikely that U.S. utilities will order additional reactors until the currently unacceptable economic risk, created by the regulatory climate and uncertain demand, is reduced. Thus it is unlikely that the private sector alone will fund major LWR design improvements. However, nuclear power will continue on its current course of expansion overseas. DOE participation is vitally needed to support the national interest in LWR technology. The report outlines R and D needs for a program to improve the safety, reliability, and economics of the present generation of plants; to develop evolutionary improved designs to be ready when needed; and to explore innovative longer-term concepts for deployment after the year 2000. The respective roles of government and the private sector are discussed

Assessment of the thorium fuel cycle in power reactors

International Nuclear Information System (INIS)

Kasten, P.R.; Homan, F.J.; Allen, E.J.

1977-01-01

A study was conducted at Oak Ridge National Laboratory to evaluate the role of thorium fuel cycles in power reactors. Three thermal reactor systems were considered: Light Water Reactors (LWRs); High-Temperature Gas-Cooled Reactors (HTGRs); and Heavy Water Reactors (HWRs) of the Canadian Deuterium Uranium Reactor (CANDU) type; most of the effort was on these systems. A summary comparing thorium and uranium fuel cycles in Fast Breeder Reactors (FBRs) was also compiled

Development Status of Accident Tolerant Fuels for Light Water Reactors in Korea

Energy Technology Data Exchange (ETDEWEB)

Yang, Jae Ho; Kim, Hyun Gil; In, Wang Kee; Kim, Weon Ju; Koo, Yang Hyum [KAERI, Daejeon (Korea, Republic of); Lee, Seung Jae [KEPCONF, Daejeon (Korea, Republic of)

2016-05-15

Research on accident tolerant fuels (ATFs) is aimed at developing innovative fuels, which can mitigate or prevent the consequences of accidents. In Korea, innovative concepts are being developed to improve fuel safety and reliability of LWRs during accident events and normal operations. ATF technologies will be developed and commercialized through a sequence of long-lead and extensive activities. The interim milestone for new fuel program is that we would be ready for an irradiation test in commercial reactor by 2021. This presentation deals with the status of ATF development in KOREA and plan to implement new fuel technology successfully in commercial nuclear power plants.

A collection of publications and articles for a light water ultra-safe plant concept

International Nuclear Information System (INIS)

Klevans, E.H.

1988-01-01

This collection contains reports titled: ''The Penn State Ultra-Safe Reactor Concept; '' ''Ultra Safe Nuclear Power; '' ''Use of the Modular Modeling System, in the Design of the Penn State Advanced Light Water Reactor; '' ''Use of the Modular Modeling System in Severe Transient Analysis of Penn State Advanced Light Water Reactor; '' ''PSU Engineers' Reactor Design May Stop a Future TMI; '' and ''The Penn State Advanced Light Water reactor Concept.''

Advanced light water reactor plant

International Nuclear Information System (INIS)

Giedraityte, Zivile

2008-01-01

For nuclear power to be competitive with the other methods of electrical power generation the economic performance should be significantly improved by increasing the time spent on line generating electricity relative to time spent off-line conducting maintenance and refueling. Maintenance includes planned actions (surveillances) and unplanned actions (corrective maintenance) to respond to component degradation or failure. A methodology is described which is used to resolve maintenance related operating cycle length barriers. Advanced light water nuclear power plant is designed with the purpose to maximize online generating time by increasing operating cycle length. (author)

Light-water reactor safety analysis codes

International Nuclear Information System (INIS)

Jackson, J.F.; Ransom, V.H.; Ybarrondo, L.J.; Liles, D.R.

1980-01-01

A brief review of the evolution of light-water reactor safety analysis codes is presented. Included is a summary comparison of the technical capabilities of major system codes. Three recent codes are described in more detail to serve as examples of currently used techniques. Example comparisons between calculated results using these codes and experimental data are given. Finally, a brief evaluation of current code capability and future development trends is presented

Utilities/industries joint study on seismic isolation systems for LWR: Part I. Experimental and analytical studies on seismic isolation systems

International Nuclear Information System (INIS)

Kato, Muneaki; Sato, Shoji; Shimomura, Issei

1989-01-01

This paper describes a joint study program on seismic isolation systems for light-water reactors (LWRs) performed by ten electric power companies, three manufacturers, and five construction companies. The fundamental response characteristics of base-isolated structures and base-isolation devices are described. Applications of a base-isolation system to LWR buildings are given. Finally, three-dimensional shaking table experiments are described

Sadhana | Indian Academy of Sciences

Indian Academy of Sciences (India)

In this study, neutronic investigation of a deuterium–tritium (DT) driven hybrid reactor using ceramic uranium fuels, namely UC, UO2 or UN under a high neutron wall load (NWL) of 10 MW/m2 at the first wall is conducted over a period of 24 months for fissile fuel breeding for light water reactors (LWRs). New substances ...

Core design concepts for high performance light water reactors

International Nuclear Information System (INIS)

Schulenberg, T.; Starflinger, J.

2007-01-01

Light water reactors operated under supercritical pressure conditions have been selected as one of the promising future reactor concepts to be studied by the Generation IV International Forum. Whereas the steam cycle of such reactors can be derived from modern fossil fired power plants, the reactor itself, and in particular the reactor core, still need to be developed. Different core design concepts shall be described here to outline the strategy. A first option for near future applications is a pressurized water reactor with 380 .deg. C core exit temperature, having a closed primary loop and achieving 2% pts. higher net efficiency and 24% higher specific turbine power than latest pressurized water reactors. More efficiency and turbine power can be gained from core exit temperatures around 500 .deg. C, which require a multi step heat up process in the core with intermediate coolant mixing, achieving up to 44% net efficiency. The paper summarizes different core and assembly design approaches which have been studied recently for such High Performance Light Water Reactors

Advanced Instrumentation, Information and Control (II&C) Research and Development Facility Buildout and Project Execution of LWRS II&C Pilot Projects 1 and 3

Energy Technology Data Exchange (ETDEWEB)

Ronald Farris; Johanna Oxstrand; Gregory Weatherby

2011-09-01

The U.S. Department of Energy (DOE) is sponsoring research, development, and deployment on light water reactor sustainability (LWRS), in which the Idaho National Laboratory (INL) is working closely with nuclear utilities to develop technologies and solutions to help ensure the safe operational life extension of current reactors. As technologies are introduced that change the operation of the plant, the LWRS pilot projects can help identify their best-advanced uses and help demonstrate the safety of these technologies. In early testing of operator performance given these emerging technologies will ensure the safety and usability of systems prior to large-scale deployment and costly verification and validation at the plant. The aim of these collaborations, demonstrations, and approaches are intended to lessen the inertia that sustains the current status quo of today's II&C systems technology, and to motivate transformational change and a shift in strategy to a long-term approach to II&C modernization that is more sustainable. Research being conducted under Pilot Project 1 regards understanding the conditions and behaviors that can be modified, either through process improvements and/or technology deployment, to improve the overall safety and efficiency of outage control at nuclear facilities. The key component of the research in this pilot project is accessing the delivery of information that will allow researchers to simulate the control room, outage control center (OCC) information, and plant status data. The simulation also allows researchers to identify areas of opportunity where plant operating status and outage activities can be analyzed to increase overall plant efficiency. For Pilot Project 3 the desire is to demonstrate the ability of technology deployment and the subsequent impact on maximizing the 'Collective Situational Awareness' of the various stakeholders in a commercial nuclear power plant. Specifically, the desire is to show positive

Tritium formation and elimination in light-water reactors

International Nuclear Information System (INIS)

Dolle, L.; Briec, M.; Miquel, P.

1976-01-01

Light-water reactors have a tritium balance which should be considered from both the working constraint and environmental pollution aspects. The formation of tritium in the primary circuit and in the fuel, the elimination and enrichment processes are considered [fr

The light water natural uranium reactor

International Nuclear Information System (INIS)

Radkowsky, A.

A new type of light water seed blanket with the seed having 20% enrichment and the blanket a special combination of elements of natural uranium and thorium, relatively close packed, but sufficient spacing for heat transfer purpose is described. The blanket would deliver approximately half the total energy for about 10,000 MWDIT, so this type of core would be just as economical or better in uranium ore consumation as present cores. (author)

Light water reactor safety research project

International Nuclear Information System (INIS)

Markoczy, G.; Aksan, S.N.; Behringer, K.; Prodan, M.; Stierli, F.; Ullrich, G.

1980-07-01

The research and development activities for the safety of Light Water Power Reactors carried out 1979 at the Swiss Federal Institute for Reactor Research are described. Considerations concerning the necessity, objectives and size of the Safety Research Project are presented, followed by a detailed discussion of the activities in the five tasks of the program, covering fracture mechanics and nondestructive testing, thermal-hydraulics, reactor noise analysis and pressure vessel steel surveillance. (Auth.)

Laser-light backscattering response to water content and proteolysis in dry-cured ham

DEFF Research Database (Denmark)

Fulladosa, E.; Rubio-Celorio, M.; Skytte, Jacob Lercke

2017-01-01

on the acquisition conditions used. Laser backscattering was influenced by both dryness and proteolysis intensity showing an average light intensity decrease of 0.2 when decreasing water content (1% weight loss) and increasing proteolysis (equivalent to one-hour enzyme action). However, a decrease of scattering area...... was only detected when the water content was decreased (618 mm(2) per 1% weight loss). Changes on scattering of light profiles were only observed when the water content changed. Although there is a good correlation between water content and LBI parameters when analysing commercial samples, proteolysis...... of laser incidence) and to analyse the laser-light backscattering changes caused by additional hot air drying and proteolysis of dry-cured ham slices. The feasibility of the technology to determine water content and proteolysis (which is related to textural characteristics) of commercial sliced dry...

HYBRID ALARM SYSTEMS: COMBINING SPATIAL ALARMS AND ALARM LISTS FOR OPTIMIZED CONTROL ROOM OPERATION

Energy Technology Data Exchange (ETDEWEB)

Ronald L. Boring; J.J. Persensky

2012-07-01

The US Department of Energy (DOE) is sponsoring research, development, and deployment on Light Water Reactor Sustainability (LWRS), in which the Idaho National Laboratory (INL) is working closely with nuclear utilities to develop technologies and solutions to help ensure the safe operational life extension of current nuclear power plants. One of the main areas of focus is control room modernization. Within control room modernization, alarm system upgrades present opportunities to meet the broader goals of the LWRS project in demonstrating the use and safety of the advanced instrumentation and control (I&C) technologies and the short-term and longer term objectives of the plant. In this paper, we review approaches for and human factors issues behind upgrading alarms in the main control room of nuclear power plants.

Realizing the world economic, environmental and non-proliferation benefits of the ALMR actinide recycle system

International Nuclear Information System (INIS)

Quinn, J.E.

1995-01-01

The Advanced Liquid Metal Reactor (ALMR) Actinide Recycle System has enormous potential to: positively impact the global environment; extend currently known energy resources; and assist developing/undeveloped countries' energy resource access. A global vision of the benefits of developing and deploying this system, including the role of highly developed countries is presented. In this vision, the ALMR System is heavily used, in conjunction with light water reactors (LWRs), in highly developed countries thus displacing fossil fuel demand in those economies, and making them more available for developing/undeveloped countries. In addition, LWRs in developing countries can further reduce global fossil fuel demand. This vision provides quality of life and standard of living improvements in these developing/undeveloped economies, which should lead to more political stability. (author)

Rapid water disinfection using vertically aligned MoS_2 nanofilms and visible light

International Nuclear Information System (INIS)

Liu, Chong; Kong, Desheng; Hsu, Po-Chun; Yuan, Hongtao; Lee, Hyun-Wook

2016-01-01

Here, solar energy is readily available in most climates and can be used for water purification. However, solar disinfection of drinking water (SODIS) mostly relies on ultraviolet light, which represents only 4% of total solar energy, and this leads to slow treatment speed. The development of new materials that can harvest visible light for water disinfection, and speed up solar water purification, is therefore highly desirable. Here, we show that few-layered vertically aligned MoS_2 (FLV-MoS_2) films can be used to harvest the whole spectrum of visible light (~ 50% of solar energy) and achieve highly efficient water disinfection. The bandgap of MoS_2 was increased from 1.3 eV to 1.55 eV by decreasing the domain size, which allowed the FLV-MoS_2 to generate reactive oxygen species (ROS) for bacterial inactivation in water. The FLV-MoS_2 showed ~15 times better log inactivation efficiency of indicator bacteria compared to bulk MoS_2, and much faster inactivation of bacteria under both visible light and sunlight illumination compared to widely used TiO_2. Moreover, by using a 5 nm copper film on top of the FLV-MoS_2 as a catalyst to facilitate electron-hole pair separation and promote the generation of ROS, the disinfection rate was further increased 6 fold. With our approach, we achieved water disinfection of >99.999% inactivation of bacteria in 20 minutes with a small amount of material (1.6 mg/L) under simulated visible light.

Rapid water disinfection using vertically aligned MoS2 nanofilms and visible light

International Nuclear Information System (INIS)

Liu, Chong; Kong, Desheng; Hsu, Po-Chun; Yuan, Hongtao; Lee, Hyun-Wook

2016-01-01

In most climates, solar energy is readily available and can be used for water purification. But, solar disinfection of drinking water mostly relies on ultraviolet light, which represents only 4% of the total solar energy, and this leads to a slow treatment speed. Therefore, the development of new materials that can harvest visible light for water disinfection, and so speed up solar water purification, is highly desirable. Here we show that few-layered vertically aligned MoS_2 (FLV-MoS_2) films can be used to harvest the whole spectrum of visible light (~50% of solar energy) and achieve highly efficient water disinfection. The bandgap of MoS_2 was increased from 1.3 to 1.55 eV by decreasing the domain size, which allowed the FLV-MoS_2 to generate reactive oxygen species (ROS) for bacterial inactivation in the water. The FLV-MoS_2 showed a ~15 times better log inactivation efficiency of the indicator bacteria compared with that of bulk MoS_2, and a much faster inactivation of bacteria under both visible light and sunlight illumination compared with the widely used TiO_2. Moreover, by using a 5 nm copper film on top of the FLV-MoS_2 as a catalyst to facilitate electron–hole pair separation and promote the generation of ROS, the disinfection rate was increased a further sixfold. Here, we achieved water disinfection of >99.999% inactivation of bacteria in 20 min with a small amount of material (1.6 mg l–1) under simulated visible light.

Development of crack growth and crack initiation test units for stress corrosion cracking examinations in high-temperature water environments under neutron irradiation (1) (Contract research)

International Nuclear Information System (INIS)

Izumo, Hironobu; Ishida, Takuya; Kawamata, Kazuo; Inoue, Shuichi; Ide, Hiroshi; Saito, Takashi; Ishitsuka, Etsuo; Chimi, Yasuhiro; Ise, Hideo; Miwa, Yukio; Ugachi, Hirokazu; Nakano, Junichi; Kaji, Yoshiyuki; Tsukada, Takashi

2009-04-01

To evaluate integrity of irradiation-assisted stress corrosion cracking (IASCC) on in-core structural materials used in light water reactors (LWRs), useful knowledge regarding IASCC has been obtained mainly by post-irradiation examinations (PIEs). In the core of commercial LWRs, however, the actual IASCC occurs under the effects of irradiation on both materials and high-temperature water environment. Therefore, it is necessary to confirm the suitability of the knowledge by PIE with comparison to IASCC behaviors during in-core SCC tests. Fundamental techniques for in-core crack growth and crack initiation tests have been developed already at the Japan Materials Testing Reactor (JMTR) of the Japan Atomic Energy Agency (JAEA). For the in-core crack growth test technique, to evaluate the effects of neutron irradiation on stainless steels irradiated to low neutron fluences, it is indispensable to develop new loading technique which is applicable to compact tension (CT) specimens with thickness of 0.5 inch (0.5T), from the viewpoint of validity based on the fracture mechanics. Based on the present technical investigation for the in-core loading technique, it is expected that a target load of 7.6 kN approximately can apply to a 0.5T-CT specimen by adopting a loading unit of a lever type instead of the previous uni-axial tension type. For the in-core crack initiation test technique, moreover, construction of a loading unit adopting linear variable differential transformers (LVDTs) has been investigated and technical issues have examined. (author)

High resolution conductometry for isotopic assay of deuterium in mixtures of heavy water and light water

International Nuclear Information System (INIS)

Ananthanarayanan, R.; Sahoo, P.; Murali, N.

2014-01-01

A PC based high resolution conductivity monitoring technique has been deployed for determination of isotopic purity of heavy water in samples containing heavy water and light water mixtures using pulsating sensor based conductivity monitoring instrument. The technique involves accurate determination of conductivities of a series of specially treated heavy water and light water mixtures of various compositions at a constant solution temperature. The shift in conductivity (Δκ), which is the difference between conductivities of composite mixture after and before the formation of a typical complex compound (boric acid–mannitol complex in this case), shows a smooth and reproducible decreasing trend with increase in percentage composition of heavy water. This relation, which is obtained by appropriate calibration, is used in the software program for direct display of isotopic purity of heavy water. The technique is examined for determination of percentage composition of heavy water in the entire range of concentration (0-100 %) with reasonable precision (relative standard deviation, RSD ≤1.5 %). About 1 mL of sample is required for each analysis and analysis is completed within a couple of minutes after pretreatment of sample. The accuracy in measurement is ≤1.75 %. (author)

Light water breeder reactor using a uranium-plutonium cycle

International Nuclear Information System (INIS)

Radkowsky, A.; Chen, R.

1990-01-01

This patent describes a light water receptor (LWR) for breeding fissile material using a uranium-plutonium cycle. It comprises: a prebreeder section having plutonium fuel containing a Pu-241 component, the prebreeder section being operable to produce enriched plutonium having an increased Pu-241 component; and a breeder section for receiving the enriched plutonium from the prebreeder section, the breeder section being operable for breeding fissile material from the enriched plutonium fuel. This patent describes a method of operating a light water nuclear reactor (LWR) for breeding fissile material using a uranium-plutonium cycle. It comprises: operating the prebreeder to produce enriched plutonium fuel having an increased Pu-241 component; fueling a breeder section with the enriched plutonium fuel to breed the fissile material

Overview of plant life extension technology development in Japan

International Nuclear Information System (INIS)

Takahashi, T.; Arai, H.; Akiyama, M.; Mishima, Y.; Okubo, T.

1993-01-01

In Japan, it is expected that the nuclear power will continue to play an important role in electric power supply. Since it is expected that the fast breeder reactor (FBR) will be introduced sometime during the first half of the 21st century, light water reactors (LWRs) will continue to play a key role some 30 to 40 years to come. For this reason, technology development projects are being implemented to further enhance light water reactor technology and thereby improve the reliability of LWRs. From this point, the Plant Life Extension (PLEX) technology development program [1-4] is entrusted by the Ministry of International Trade and Industry to the Japan Power Engineering and Inspection Corporation (JAPEIC). This program is an 11-year plan which started in 1985. The objectives are to extend the service lives of existing LWRs to increase the energy generated by these plants during their lifetime, and to reduce the lifetime generating costs. In this report, we will present our project overview and recent activities with respect to extensive verification tests on component material behavior. The JAPEIC PLEX project is divided into 3 phases. Phase I is the feasibility study. Phase 2 involves the verification tests and the evaluation of life extension technologies. The overall evaluation of the project will be conducted in Phase 3. The feasibility study of Phase I has been completed in fiscal year 1985 and 1986. In Phase I, the important components (the components and structures that are likely to govern the lives of nuclear power plants) have been selected. (author)

Technology Implementation Plan. Fully Ceramic Microencapsulated Fuel for Commercial Light Water Reactor Application

International Nuclear Information System (INIS)

Snead, Lance Lewis; Terrani, Kurt A.; Powers, Jeffrey J.; Worrall, Andrew; Robb, Kevin R.; Snead, Mary A.

2015-01-01

This report is an overview of the implementation plan for ORNL's fully ceramic microencapsulated (FCM) light water reactor fuel. The fully ceramic microencapsulated fuel consists of tristructural isotropic (TRISO) particles embedded inside a fully dense SiC matrix and is intended for utilization in commercial light water reactor application.

Technology Implementation Plan. Fully Ceramic Microencapsulated Fuel for Commercial Light Water Reactor Application

Energy Technology Data Exchange (ETDEWEB)

Snead, Lance Lewis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Worrall, Andrew [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Snead, Mary A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-04-01

This report is an overview of the implementation plan for ORNL's fully ceramic microencapsulated (FCM) light water reactor fuel. The fully ceramic microencapsulated fuel consists of tristructural isotropic (TRISO) particles embedded inside a fully dense SiC matrix and is intended for utilization in commercial light water reactor application.

Steam explosions in light water reactors

International Nuclear Information System (INIS)

1981-01-01

The report deals with a postulated accident caused by molten fuel falling into the lower plenum of the containment of a reactor. The analysis which is presented in the report shows that the thermal energy released in the resulting steam explosion is not enough to destroy the pressure vessel or the containment. The report was prepared for the Swedish Governmental Committee on steam explosion in light water reactors. It includes statements issued by internationally well-known specialists. (G.B.)

Programme of research and development on plutonium recycling in light-water reactors

International Nuclear Information System (INIS)

1979-01-01

This is the third annual progress report concerning the programme on plutonium recycling in light-water reactors (indirect action) of the Commission of the European Communities. It covers the year 1978 and follows the annual reports for 1977 (EUR 6002 EN) and 1976 (EUR 5780). The preliminary results obtained under the 1975-79 programme indicate that: (a) assuming that plutonium recycling in light-water reactors is industrially developed by the end of the century, the foreseeable radiological impact on both workers and the general public can be maintained within the limits of current radiation protection standards; (b) on the whole, there is a good knowledge and mastery of the specific aspects involved in the plutonium recycling in light-water reactors and in particular they indicate that plutonium fuels have a similar behaviour to uranium fuels

Evaluation of improved light water reactor core designs. Final progress report, September 1979. LWRCD-20

International Nuclear Information System (INIS)

1979-01-01

The work conducted under this research project has developed information which supports in all respects the U.S. position evolved under the NASAP/INFCE programs with respect to the near and intermediate term potential for ore conservation in LWRs on the once-through fuel cycle. Moreover, in the even longer term, it has been confirmed that contention by Edlund and others that tight-pitch Pu/UO 2 PWR cores can achieve conversion ratios which may allow these reactors to provide a competitive energy source far into the ore-scarce post-2000 era

Self-propagating solar light reduction of graphite oxide in water

Energy Technology Data Exchange (ETDEWEB)

Todorova, N.; Giannakopoulou, T.; Boukos, N.; Vermisoglou, E. [Institute of Nanoscience and Nanotechnology, NCSR “Demokritos”, 153 41 Attikis (Greece); Lekakou, C. [Division of Mechanical, Medical, and Aerospace Engineering, Faculty of Engineering and Physical Sciences, University of Surrey, Guildford (United Kingdom); Trapalis, C., E-mail: c.trapalis@inn.demokritos.gr [Institute of Nanoscience and Nanotechnology, NCSR “Demokritos”, 153 41 Attikis (Greece)

2017-01-01

Highlights: • Graphite oxide was partially reduced by solar light irradiation in water media. • No addition of catalysts nor reductive agent were used for the reduction. • Specific capacitance increased stepwise with increase of irradiation time. • Self-propagating reduction of graphene oxide by solar light is suggested. - Abstract: Graphite Oxide (GtO) is commonly used as an intermediate material for preparation of graphene in the form of reduced graphene oxide (rGO). Being a semiconductor with tunable band gap rGO is often coupled with various photocatalysts to enhance their visible light activity. The behavior of such rGO-based composites could be affected after prolonged exposure to solar light. In the present work, the alteration of the GtO properties under solar light irradiation is investigated. Water dispersions of GtO manufactured by oxidation of natural graphite via Hummers method were irradiated into solar light simulator for different periods of time without addition of catalysts or reductive agent. The FT-IR analysis of the treated dispersions revealed gradual reduction of the GtO with the increase of the irradiation time. The XRD, FT-IR and XPS analyses of the obtained solid materials confirmed the transition of GtO to rGO under solar light irradiation. The reduction of the GtO was also manifested by the CV measurements that revealed stepwise increase of the specific capacitance connected with the restoration of the sp{sup 2} domains. Photothermal self-propagating reduction of graphene oxide in aqueous media under solar light irradiation is suggested as a possible mechanism. The self-photoreduction of GtO utilizing solar light provides a green, sustainable route towards preparation of reduced graphene oxide. However, the instability of the GtO and partially reduced GO under irradiation should be considered when choosing the field of its application.

Multiphase flow in ex-vessel coolability: development of an innovative concept

International Nuclear Information System (INIS)

Corradini, Michael L.

2006-01-01

The interaction and mixing of high-temperature melt and water is the important technical issue in the safety assessment of water-cooled reactors to achieve ultimate core coolability. For specific Advanced Light Water Reactor (ALWR) designs, deliberate mixing of the core-melt and water is being considered as a mitigative measure, to assure ex-vessel core coolability. The paper provides the background of past experiments as well as key fundamentals that are needed for melt-water interfacial transport phenomena, thus enabling the development of innovative safety technologies for advanced LWRs that will assure ex-vessel core coolability

Analysis of an accelerator-driven subcritical light water reactor

International Nuclear Information System (INIS)

Kruijf, W.J.M. de; Wakker, P.H.; Wetering, T.F.H. van de; Verkooijen, A.H.M.

1997-01-01

An analysis of the basic characteristics of an accelerator-driven light water reactor has been made. The waste in the nuclear fuel cycle is considerably less than in the light water reactor open fuel cycle. This is mainly caused by the use of equilibrium nuclear fuel in the reactor. The accelerator enables the use of a fuel composition with infinite multiplication factor k ∞ < 1. The main problem of the use of this type of fuel is the strongly peaked flux distribution in the reactor core. A simple analytical model shows that a large core is needed with a high peak power factor in order to generate net electric energy. The fuel in the outer regions of the reactor core is used very poorly. 7 refs., 4 figs., 1 tab

Penn State advanced light water reactor concept

International Nuclear Information System (INIS)

Borkowski, J.A.; Smith, K.A.; Edwards, R.M.; Robinson, G.E.; Schultz, M.A.; Klevans, E.H.

1987-01-01

The accident at Three Mile Island heightened concerns over the safety of nuclear power. In response to these concerns, a research group at the Pennsylvania State University (Penn State) undertook the conceptual design of an advanced light water reactor (ALWR) under sponsorship of the US Dept. of Energy (DOE). The design builds on the literally hundreds of years worth of experience with light water reactor technology. The concept is a reconfigured pressurized water reactor (PWR) with the capability of being shut down to a safe condition simply by removing all ac power, both off-site and on-site. Using additional passively activated heat sinks and replacing the pressurizer with a pressurizing pump system, the concept essentially eliminates the concerns of core damage associated with a total station blackout. Evaluation of the Penn State ALWR concept has been conducted using the EPRI Modular Modeling System (MMS). Results show that a superior response to normal operating transients can be achieved in comparison to the response with a conventional PWR pressurizer. The DOE-sponsored Penn State ALWR concept has evolved into a significant reconfiguration of a PWR leading to enhanced safety characteristics. The reconfiguration has touched a number of areas in overall plant design including a shutdown turbine in the secondary system, additional passively activated heat sinks, a unique primary side pressurizing concept, a low pressure cleanup system, reactor building layout, and a low power density core design

Light-water-reactor hydrogen manual

International Nuclear Information System (INIS)

Camp, A.L.; Cummings, J.C.; Sherman, M.P.; Kupiec, C.F.; Healy, R.J.; Caplan, J.S.; Sandhop, J.R.; Saunders, J.H.

1983-06-01

A manual concerning the behavior of hydrogen in light water reactors has been prepared. Both normal operations and accident situations are addressed. Topics considered include hydrogen generation, transport and mixing, detection, and combustion, and mitigation. Basic physical and chemical phenomena are described, and plant-specific examples are provided where appropriate. A wide variety of readers, including operators, designers, and NRC staff, will find parts of this manual useful. Different sections are written at different levels, according to the most likely audience. The manual is not intended to provide specific plant procedures, but rather, to provide general guidance that may assist in the development of such procedures

Burning actinides in very hard spectrum reactors

International Nuclear Information System (INIS)

Robinson, A.H.; Shirley, G.W.; Prichard, A.W.; Trapp, T.J.

1978-01-01

The major unresolved problem in the nuclear industry is the ultimate disposition of the waste products of light water reactors. The study demonstrates the feasibility of designing a very hard spectrum actinide burner reactor (ABR). A 1100 MW/sub t/ ABR design fueled entirely with actinides reprocessed from light water reactor (LWR) wastes is proposed as both an ultimate disposal mechanism for actinides and a means of concurrently producing usable power. Actinides from discharged ABR fuel are recycled to the ABR while fission products are routed to a permanent repository. As an integral part of a large energy park, each such ABR would dispose of the waste actinides from 2 LWRs

Modification of MELCOR for severe accident analysis of candidate accident tolerant cladding materials

Energy Technology Data Exchange (ETDEWEB)

Merrill, Brad J., E-mail: brad.merrill@inl.gov; Bragg-Sitton, Shannon M., E-mail: shannon.bragg-sitton@inl.gov; Humrickhouse, Paul W., E-mail: paul.humrickhouse@inl.gov

2017-04-15

Highlights: • Accident tolerant fuels (ATF) systems are currently under development for LWRs. • Many performance analysis tools are specifically developed for UO{sub 2}–Zr alloy fuel. • Modifications were made to the MELCOR code for candidate ATF cladding. • Preliminary analysis results for SiC and FeCrAl cladding concepts are presented. - Abstract: A number of materials are currently under development as candidate accident tolerant fuel and cladding for application in the current fleet of commercial light water reactors (LWRs). The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Enhancing the accident tolerance of light water reactors became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal for the development of accident tolerant fuel (ATF) systems for LWRs is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. Designed for use in the current fleet of commercial LWRs, or in reactor concepts with design certifications (GEN-III+), to achieve their goal enhanced ATF must endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system, while maintaining or improving performance during normal operation. Many available nuclear fuel performance analysis tools are specifically developed for the current UO{sub 2}–Zirconium alloy fuel system. The MELCOR severe-accident analysis code, under development at the Sandia National Laboratory in New Mexico (SNL-NM) for the US Nuclear Regulatory Commission (NRC), is one of these tools. This paper describes modifications

Light Refraction by Water as a Rationale for the Poggendorff Illusion

DEFF Research Database (Denmark)

Bozhevolnyi, Sergey I.

2017-01-01

The Poggendorff illusion in its classical form of parallel lines interrupting a transversal is viewed from the perspective of being related to the everyday experience of observing the light refraction in water. It is argued that if one considers a transversal to be a light ray in air and the para...

Beyond the light water reactor

International Nuclear Information System (INIS)

Nero, A.V. Jr.

1980-01-01

One of the strong interests in examining alternative nuclear fuel cycles is to search for schemes that are more efficient than LWRs in their use of uranium, but that do not carry the additional proliferation hazard associated with widespread plutonium utilization. One possibility is to improve the uranium efficiency of current reactor types by other means than recycling. A second possibility, offering greater potential for improvement, is to utilize thorium-uranium fuel cycles in which uranium-233 is denatured by the addition of uranium-238, making enrichment necessary to yield weapons-usable material. The bulk of the reactor's fuel material would be thorium-232, so that most of the fissile material produced would be uranium-233. It is important to recognize that these two possibilities - once through improvements and denatured thorium-uranium - could be introduced sequentially in reactor types that are currently in use. Fuel cycles may change over time, but it is equally significant from the point of view of non-proliferation that they will also vary from place to place and, most importantly, from country to country. The author argues that alternative nuclear power systems and a slower growth may affect the diversion of nuclear materials to weapons. A real question, though, is whether we have time to explore the possibilities. It has become apparent that predictions made of the growth rate for nuclear power were too high. The 1000 large power plants the US was to have by the year 2000 have been reduced to fewer than 300. This reduces the pressure, resulting from our limited uranium resources, to push the LMFBR. Extra time gives us a chance to examine the possibilities

Separation setup for the light water detritiation process in the water-hydrogen system based on the membrane contact devices

International Nuclear Information System (INIS)

Rozenkevich, M. B.; Rastunova, I. L.; Prokunin, S. V.

2008-01-01

Detritiation of light water wastes down to a level permissible to discharge into the environment while simultaneously concentrating tritium to decrease amount of waste being buried is a constant problem. The laboratory setup for the light water detritiation process is presented. The separation column consists of 10 horizontally arranged perfluorosulphonic acid Nafion-type membrane contact devises and platinum catalyst (RCTU-3SM). Each contact device has 42.3 cm 2 of the membrane and 10 cm 3 of the catalyst. The column is washed by tritium free light water (L H2O ) and the tritium-containing flow (F HTO ) feeds the electrolyser at λ = G H2 /L H2O = 2. A separation factor of 66 is noted with the device at 336 K and 0.145 MPa. (authors)

Characteristics of ultraviolet light and radicals formed by pulsed discharge in water

Science.gov (United States)

Sun, Bing; Kunitomo, Shinta; Igarashi, Chiaki

2006-09-01

In this investigation, the ultraviolet light characteristics and OH radical properties produced by a pulsed discharge in water were studied. For the plate-rod reactor, it was found that the ultraviolet light energy has a 3.2% total energy injected into the reactor. The ultraviolet light changed with the peak voltage and electrode distance. UV characteristics in tap water and the distilled water are given. The intensity of the OH radicals was the highest for the 40 mm electrode distance reactor. In addition, the properties of hydrogen peroxide and ozone were also studied under arc discharge conditions. It was found that the OH radicals were in the ground state and the excited state when a pulsed arc discharge was used. The ozone was produced by the arc discharge even if the oxygen gas is not bubbled into the reactor. The ozone concentration produces a maximum value with treatment time.

Characteristics of ultraviolet light and radicals formed by pulsed discharge in water

Energy Technology Data Exchange (ETDEWEB)

Sun Bing [Dalian Maritime University, College of Environment, 1st Linghai Road, Dalian (China); Kunitomo, Shinta [Ebara Corporation, 1-6-27, Konan, Minato-ku 108-8480 (Japan); Igarashi, Chiaki [Ebara Research Co. Ltd, 2-1, Honfujisawa 4-chome, Fujisawa 251-8502 (Japan)

2006-09-07

In this investigation, the ultraviolet light characteristics and OH radical properties produced by a pulsed discharge in water were studied. For the plate-rod reactor, it was found that the ultraviolet light energy has a 3.2% total energy injected into the reactor. The ultraviolet light changed with the peak voltage and electrode distance. UV characteristics in tap water and the distilled water are given. The intensity of the OH radicals was the highest for the 40 mm electrode distance reactor. In addition, the properties of hydrogen peroxide and ozone were also studied under arc discharge conditions. It was found that the OH radicals were in the ground state and the excited state when a pulsed arc discharge was used. The ozone was produced by the arc discharge even if the oxygen gas is not bubbled into the reactor. The ozone concentration produces a maximum value with treatment time.

Effect of cold water and inverse lighting on growth performance of broiler chickens under extreme heat stress.

Science.gov (United States)

Park, Sang-oh; Park, Byung-sung; Hwangbo, Jong

2015-07-01

The present study was carried out to investigate the effect of provision of extreme heat stress diet (EHD), inverse lighting, cold water on growth performance of broiler chickens exposed to extreme heat stress. The chickens were divided into four treatment groups, (T1, T2, T3, T4) as given below: Ti (EHD 1, 10:00-19:00 dark, 19:00-10:00 light, cool water 9 degrees C); T2 (EHD 2, 10:00-19:00 dark, 19:00-10:00 light, cool water 9 degrees C); T3 (EHD 1, 09:00-18:00 dark, 18:00-09:00 light, cool water 141C); T4 (EHD 2, 09:00-18:00 dark, 18:00-09:00 light, cool water 14 degrees C. EHD 1 contained soybean oil, molasses, methionine and lysine; EHD 2 contained the same ingredients as EHD 1 with addition of vitamin C. Groups T1 and T2 were given cooler water than the othertwo groups, and displayed higher body weight increase and diet intake as compared to T3 and T4 (pstress diet, inverse lighting (10:00-19:00 dark, 19:00-10:00 light) with cold water at 9 degrees C under extreme heat stress could enhance growth performance of broiler chickens.

Advanced Instrumentation, Information and Control (II and C) Research and Development Facility Buildout and Project Execution of LWRS II and C Pilot Projects 1 and 3

International Nuclear Information System (INIS)

Farris, Ronald; Oxstrand, Johanna; Weatherby, Gregory

2011-01-01

The U.S. Department of Energy (DOE) is sponsoring research, development, and deployment on light water reactor sustainability (LWRS), in which the Idaho National Laboratory (INL) is working closely with nuclear utilities to develop technologies and solutions to help ensure the safe operational life extension of current reactors. As technologies are introduced that change the operation of the plant, the LWRS pilot projects can help identify their best-advanced uses and help demonstrate the safety of these technologies. In early testing of operator performance given these emerging technologies will ensure the safety and usability of systems prior to large-scale deployment and costly verification and validation at the plant. The aim of these collaborations, demonstrations, and approaches are intended to lessen the inertia that sustains the current status quo of today's II and C systems technology, and to motivate transformational change and a shift in strategy to a long-term approach to II and C modernization that is more sustainable. Research being conducted under Pilot Project 1 regards understanding the conditions and behaviors that can be modified, either through process improvements and/or technology deployment, to improve the overall safety and efficiency of outage control at nuclear facilities. The key component of the research in this pilot project is accessing the delivery of information that will allow researchers to simulate the control room, outage control center (OCC) information, and plant status data. The simulation also allows researchers to identify areas of opportunity where plant operating status and outage activities can be analyzed to increase overall plant efficiency. For Pilot Project 3 the desire is to demonstrate the ability of technology deployment and the subsequent impact on maximizing the 'Collective Situational Awareness' of the various stakeholders in a commercial nuclear power plant. Specifically, the desire is to show positive results

Development of irradiation technique with satured temperature capsule in the JMTR

International Nuclear Information System (INIS)

Ohtaka, Kimihiro

1999-01-01

The irradiation assisted stress corrosion cracking (IASCC) of in-core structural materials caused by the simultaneous effects of neutron irradiation and high temperature water environments has been pointed out as one of the major concerns not only for the light water reactors (LWRs) but also for the water-cooled fusion reactor, i.e,. ITER. The IASCC of the austenitic stainless steels or nickel base alloys has been studied for more than ten years under international efforts in the various projects for the plant life assessment and extension of LWRs. However its mechanism has not been clarified yet in spite of the extensive post-irradiation examinations. Under this situation, it is desired to perform irradiation tests under specially controlled conditions so that the effect of irradiation and high temperature water can be separately evaluated. In the Japan Materials Testing Reactor (JMTR), irradiation technique with the saturation temperature capsule (SATCAP) was developed for irradiation of the materials in the water with high, but constant, temperature and applied to study the IASCC. The capability of the SATCAP was improved by enhancing the temperature controllability to irradiate materials even in a low gamma region in the JMTR core. The performance tests of the improved SATCAP carried out in the JMTR have proven its capabilities. Based on experiences of the SATCAP, preliminary design study for the upgraded in-pile test facility are now underway in the JMTR. The test facility has a new test loop to achieve irradiate test simulated water environment of LWRs. The design, test results of the SATCAP and the design study of upgraded in-pile test facility are described in this paper

Modeling defect and fission gas properties in U-Si fuels

Energy Technology Data Exchange (ETDEWEB)

Andersson, Anders David Ragnar [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanek, Christopher Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Noordhoek, Mark [Univ. of South Carolina, Columbia, SC (United States); Besmann, Theodore [Univ. of South Carolina, Columbia, SC (United States); Middleburgh, Simon C. [Westinghouse Electric Sweden, Vasteras (Sweden); Lahoda, E. J. [Westinghouse Electric Company LLC, Cranberry Woods, PA (United States); Chernatynskiy, Aleksandr [Missouri University of Science and Technology; Grimes, Robin W. [Imperial College, London (United Kingdom)

2017-04-27

Uranium silicides, in particular U₃Si₂, are being explored as an advanced nuclear fuel with increased accident tolerance as well as competitive economics compared to the baseline UO₂ fuel. They benefit from high thermal conductivity (metallic) compared to UO₂ fuel (insulator or semi-conductor) used in current Light Water Reactors (LWRs). The U-Si fuels also have higher fissile density. In order to perform meaningful engineering scale nuclear fuel performance simulations, the material properties of the fuel, including the response to irradiation environments, must be known. Unfortunately, the data available for USi fuels are rather limited, in particular for the temperature range where LWRs would operate. The ATF HIP is using multi-scale modeling and simulations to address this knowledge gap.

Modeling defect and fission gas properties in U-Si fuels

Energy Technology Data Exchange (ETDEWEB)

Andersson, Anders David Ragnar [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanek, Christopher Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Noordhoek, Mark J. [Univ. of South Carolina, Columbia, SC (United States); Besmann, Theodore M. [Univ. of South Carolina, Columbia, SC (United States); Middleburgh, Simon C. [Westinghouse Electric Sweden, Vasteras (Sweden); Lahoda, E. J. [Westinghouse Electric Company LLC, Cranberry Woods, PA (United States); Chernatynskiy, Aleksandr [Missouri Univ. of Science and Technology, Rolla, MO (United States); Grimes, Robin W. [Imperial College, London (United Kingdom)

2017-04-14

Uranium silicides, in particular U₃Si₂, are being explored as an advanced nuclear fuel with increased accident tolerance as well as competitive economics compared to the baseline UO2 fuel. They benefit from high thermal conductivity (metallic) compared to UO₂ fuel (insulator or semi-conductor) used in current Light Water Reactors (LWRs). The U-Si fuels also have higher fissile density. In order to perform meaningful engineering scale nuclear fuel performance simulations, the material properties of the fuel, including the response to irradiation environments, must be known. Unfortunately, the data available for USi fuels are rather limited, in particular for the temperature range where LWRs would operate. The ATF HIP is using multi-scale modeling and simulations to address this knowledge gap.

Bioassay using the water soluble fraction of a Nigerian Light Crude ...

African Journals Online (AJOL)

Summary: A 96-hour bioassay was conducted using the water soluble fraction of a Nigerian light crude oil sample on Clarias gariepinus fingerlings. 0, 2.5, 5.0, 7.5 and 10 mls of water soluble fractions (WSF) of the oil were added to 1000 litres of de-chlorinated tap water to form 0, 25, 50 , 75 and 100 parts per million ...

Status of advanced light water cooled reactor designs 1996

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-09-01

The present report, which is significantly more comprehensive than the previously one, addresses the rationale and basic motivations that lead to a continuing development of nuclear technology, provides an overview of the world status of current LWRs, describes the present market situations, and identifies desired characteristics for future plants. The report also provides a detailed description of utility requirements that largely govern today`s nuclear development efforts, the situation with regard to enhanced safety objectives, a country wise description of the development activities, and a technical description of the various reactor designs in a consistent format. The reactor designs are presented in two categories: (1) evolutionary concepts that are expected to be commercially available soon; and (2) innovative designs. The report addresses the main technical characteristics of each concept without assessing or evaluating them from a particular point of view (e.g. safety or economics). Additionally, the report identifies basic reference documents that can provide further information for detailed evaluations. The report closes with an outlook on future energy policy developments.

Status of advanced light water cooled reactor designs 1996

International Nuclear Information System (INIS)

1997-09-01

The present report, which is significantly more comprehensive than the previously one, addresses the rationale and basic motivations that lead to a continuing development of nuclear technology, provides an overview of the world status of current LWRs, describes the present market situations, and identifies desired characteristics for future plants. The report also provides a detailed description of utility requirements that largely govern today's nuclear development efforts, the situation with regard to enhanced safety objectives, a country wise description of the development activities, and a technical description of the various reactor designs in a consistent format. The reactor designs are presented in two categories: (1) evolutionary concepts that are expected to be commercially available soon; and (2) innovative designs. The report addresses the main technical characteristics of each concept without assessing or evaluating them from a particular point of view (e.g. safety or economics). Additionally, the report identifies basic reference documents that can provide further information for detailed evaluations. The report closes with an outlook on future energy policy developments

Study of the light emitted in the moderation of a heavy-water pile

International Nuclear Information System (INIS)

Breton, D.

1958-06-01

During the running of a reactor which uses water as a neutron moderator, a bluish light is seen to appear inside the liquid. A detailed study of this radiation, undertaken on the Fontenay-aux-Roses pile, has shown that the spectrum is identical with that which characterises the light produced by the Cerenkov effect. The light intensity as a function of the pile power grows exponentially as a function of time when the pile diverges, with a lifetime equal to that of the rise in power. An examination of the various particles present in the pile has led to the conclusion that only electrons with an energy greater than 260 keV con produce the Cerenkov light. The light source thus produced is about 2.10 6 photons/cm 2 of water, when the pile power equals 1 watt. (author) [fr

Mechanical design of a light water breeder reactor

International Nuclear Information System (INIS)

Fauth, W.L. Jr.; Jones, D.S.; Kolsun, G.J.; Erbes, J.G.; Brennan, J.J.; Weissburg, J.A.; Sharbaugh, J.E.

1976-01-01

In a light water reactor system using the thorium-232--uranium-233 fuel system in a seed-blanket modular core configuration having the modules arranged in a symmetrical array surrounded by a reflector blanket region, the seed regions are disposed for a longitudinal movement between the fixed or stationary blanket region which surrounds each seed region. Control of the reactor is obtained by moving the inner seed region thus changing the geometry of the reactor, and thereby changing the leakage of neutrons from the relatively small seed region into the blanket region. The mechanical design of the Light Water Breeder Reactor (LWBR) core includes means for axially positioning of movable fuel assemblies to achieve the neutron economy required of a breeder reactor, a structure necessary to adequately support the fuel modules without imposing penalties on the breeding capability, a structure necessary to support fuel rods in a closely packed array and a structure necessary to direct and control the flow of coolant to regions in the core in accordance with the heat transfer requirements. 4 claims, 24 drawing figures

Assessment of the linear power level in fuel rods irradiated in the CALLISTO loop in the high flux materials testing reactor BR2

International Nuclear Information System (INIS)

Malambu, E.; Raedt, Ch. de; Weber, M.

1999-01-01

The pressurized light-water-cooled testing facility CALLISTO was designed to test the behaviour of advanced fuel rods (UO 2 or MOX, possibly with burnable poisons) under conditions representative of actual LWRs up to high burn-up rates. The accurate determination of the fission powers in each of the nine rods, and hence of the burn-up values, is carried out according to a rather elaborate procedure. (author)

Next Generation Nuclear Plant Project 2009 Status Report

Energy Technology Data Exchange (ETDEWEB)

Larry Demick; Jim Kinsey; Keith Perry; Dave Petti

2010-05-01

The mission of the NGNP Project is to broaden the environmental and economic benefits of nuclear energy technology to the United States and other economies by demonstrating its applicability to market sectors not served by light water reactors (LWRs). Those markets typically use fossil fuels to fulfill their energy needs, and high temperature gas-cooled reactors (HTGRs) like the NGNP can reduce this dependence and the resulting carbon footprint.

Light requirements of water lobelia (Lobelia dortmanna L.

Directory of Open Access Journals (Sweden)

Borowiak Dariusz

2017-12-01

Full Text Available Maximum depth of colonization (zC and total area covered by a population of Lobelia dortmanna, as well as underwater light regime were studied in 25 soft water lobelia lakes in north-western Poland. Variations in underwater light conditions among the lakes were described by Secchi disc depths (zSD, and by attenuation coefficients of irradiance within photosynthetically active radiation range (Kd,PAR, and euphotic zone depths (zEU derived from photometric measurements conducted twice a year (in midspring and midsummer during the period 2014–2015. Maximum depth of colonization of water lobelia ranged from 0.1 to 2.2 m (median zC = 0.8 m; mean zC = 1.0 m. Nine lakes showed the relative coverage of the littoral zone (RCLZ by L. dortmanna to be greater than the mean value, which was 4.8%. Studies showed that light requirements of water lobelia increase when the maximum depth of colonization also increases. This pattern could be partially related to the greater energy needs of deeper growing individuals due to enlarged seed production and their incubation, and for the creation of much heavier inflorescences. Assessment of the light requirements of L. dortmanna along the depth gradient indicates that relative irradiance (percentage of subsurface irradiance of PAR should be at the level of: (i 47–50% (annual total of quantum irradiance 3083–3280 mol m−2 yr−2 for plants growing within a depth range of 2.0–2.5 m; (ii 44–47% (2886–3083 mol m−2yr−1 for plants growing within a depth range of 1.5–2.0 m; (iii 41–44% (2690–2886 mol m−2yr−2 for plants growing within a depth range of 1.0–1.5 m; and (iv 34–41% (2230–2690 mol m−1 yr−1 for those growing in the littoral zone at a depth of between 0.5 and 1.0 m. In average conditions in the Pomeranian lakes, the maximum depth of colonization by L. dortmanna accounts for approximately a third of the Secchi disc depth and a fifth of the depth of the euphotic zone with irradiance

Aging management of major light water reactor components

International Nuclear Information System (INIS)

Shah, V.N.; Sinha, U.P.; Ware, A.G.

1992-01-01

Review of technical literature and field experience has identified stress corrosion cracking as one of the major degradation mechanisms for the major light water reactor components. Three of the stress corrosion cracking mechanisms of current concern are (a) primary water stress corrosion cracking (PWSCC) in pressurized water reactors, and (b) intergranular stress corrosion cracking (IGSCC) and (c) irradiation-assisted stress corrosion cracking (IASCC) in boiling water reactors. Effective aging management of stress corrosion cracking mechanisms includes evaluation of interactions between design, materials, stressors, and environment; identification and ranking of susceptible sites; reliable inspection of any damage; assessment of damage rate; mitigation of damage; and repair and replacement using corrosion-resistant materials. Management of PWSCC includes use of lower operating temperatures, reduction in residual tensile stresses, development of reliable inspection techniques, and use of Alloy 690 as replacement material. Management of IGSCC of nozzle and attachment welds includes use of Alloy 82 as weld material, and potential use of hydrogen water chemistry. Management of IASCC also includes potential use of hydrogen water chemistry

IAEA activities in technology development for advanced water-cooled nuclear power plants

International Nuclear Information System (INIS)

Juhn, Poong Eil; Kupitz, Juergen; Cleveland, John; Lyon, Robert; Park, Je Won

2003-01-01

As part of its Nuclear Power Programme, the IAEA conducts activities that support international information exchange, co-operative research and technology assessments and advancements with the goal of improving the reliability, safety and economics of advanced water-cooled nuclear power plants. These activities are conducted based on the advice, and with the support, of the IAEA Department of Nuclear Energy's Technical Working Groups on Advanced Technologies for Light Water Reactors (LWRs) and Heavy Water Reactors (HWRs). Assessments of projected electricity generation costs for new nuclear plants have shown that design organizations are challenged to develop advanced designs with lower capital costs and short construction times, and sizes, including not only large evolutionary plants but also small and medium size plants, appropriate to grid capacity and owner financial investment capability. To achieve competitive costs, both proven means and new approaches should be implemented. The IAEA conducts activities in technology development that support achievement of improved economics of water-cooled nuclear power plants (NPPs). These include fostering information sharing and cooperative research in thermo-hydraulics code validation; examination of natural circulation phenomena, modelling and the reliability of passive systems that utilize natural circulation; establishment of a thermo-physical properties data base; improved inspection and diagnostic techniques for pressure tubes of HWRs; and collection and balanced reporting from recent construction and commissioning experiences with evolutionary water-cooled NPPs. The IAEA also periodically publishes Status Reports on global development of advanced designs. (author)

Impact of different moderator ratios with light and heavy water cooled reactors in equilibrium states

International Nuclear Information System (INIS)

Permana, Sidik; Takaki, Naoyuki; Sekimoto, Hiroshi

2006-01-01

As an issue of sustainable development in the world, energy sustainability using nuclear energy may be possible using several different ways such as increasing breeding capability of the reactors and optimizing the fuel utilization using spent fuel after reprocessing as well as exploring additional nuclear resources from sea water. In this present study the characteristics of light and heavy water cooled reactors for different moderator ratios in equilibrium states have been investigated. The moderator to fuel ratio (MFR) is varied from 0.1 to 4.0. Four fuel cycle schemes are evaluated in order to investigate the effect of heavy metal (HM) recycling. A calculation method for determining the required uranium enrichment for criticality of the systems has been developed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of SRAC 2000 code using nuclear data library from the JENDL 3.2. The results show a thermal spectrum peak appears for light water coolant and no thermal peak for heavy water coolant along the MFR (0.1 ≤ MFR ≤ 4.0). The plutonium quality can be reduced effectively by increasing the MFR and number of recycled HM. Considering the effect of increasing number of recycled HM; it is also effective to reduce the uranium utilization and to increase the conversion ratio. trans-Plutonium production such as americium (Am) and curium (Cm) productions are smaller for heavy water coolant than light water coolant. The light water coolant shows the feasibility of breeding when HM is recycled with reducing the MFR. Wider feasible area of breeding has been obtained when light water coolant is replaced by heavy water coolant

Assessment of predictive capability of REFLA/TRAC code for large break LOCA transient in PWR using LOFT L2-5 test data

International Nuclear Information System (INIS)

Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio

1994-03-01

The REFLA/TRAC code is a best estimate code developed at Japan Atomic Energy Research Institute (JAERI) to provide advanced predictions of thermal hydraulic transient in light water reactors (LWRs). The REFLA/TRAC code uses the TRAC-PF1/MOD1 code as the framework of the code. The REFLA/TRAC code is expected to be used for the calibration of licensing codes, accident analysis, accident simulation of LWRs, and design of advanced LWRs. Several models have been implemented to the TRAC-PF1/MOD1 code at JAERI including reflood model, condensation model, interfacial and wall friction models, etc. These models have been verified using data from various separate effect tests. This report describes an assessment result of the REFLA/TRAC code, which was performed to assess the predictive capability for integral system behavior under large break loss of coolant accident (LBLOCA) using data from the LOFT L2-5 test. The assessment calculation confirmed that the REFLA/TRAC code can predict break mass flow rate, emergency core cooling water bypass and clad temperature excellently in the LOFT L2-5 test. The CPU time of the REFLA/TRAC code was about 1/3 of the TRAC-PF1/MOD1 code. The REFLA/TRAC code can perform stable and fast simulation of thermal hydraulic behavior in PWR LBLOCA with enough accuracy for practical use. (author)

The lantern shark's light switch: turning shallow water crypsis into midwater camouflage

Science.gov (United States)

Claes, Julien M.; Mallefet, Jérôme

2010-01-01

Bioluminescence is a common feature in the permanent darkness of the deep-sea. In fishes, light is emitted by organs containing either photogenic cells (intrinsic photophores), which are under direct nervous control, or symbiotic luminous bacteria (symbiotic photophores), whose light is controlled by secondary means such as mechanical occlusion or physiological suppression. The intrinsic photophores of the lantern shark Etmopterus spinax were recently shown as an exception to this rule since they appear to be under hormonal control. Here, we show that hormones operate what amounts to a unique light switch, by acting on a chromatophore iris, which regulates light emission by pigment translocation. This result strongly suggests that this shark's luminescence control originates from the mechanism for physiological colour change found in shallow water sharks that also involves hormonally controlled chromatophores: the lantern shark would have turned the initial shallow water crypsis mechanism into a midwater luminous camouflage, more efficient in the deep-sea environment. PMID:20410033

Is light water reactor technology sustainable?

International Nuclear Information System (INIS)

Rothwell, G.; Van der Zwaan, B.

2001-01-01

This paper proposes criteria for determining ''intermediate sustainability'' over a 500-year horizon. We apply these criteria to Light Water Reactor (LWR) technology and the LWR industry. We conclude that LWR technology does not violate intermediate sustainability criteria for (1) environmental externalities, (2) worker and public health and safety, or (3) accidental radioactive release. However, it does not meet criteria to (1) efficiently use depleted uranium and (2) avoid uranium enrichment technologies that can lead to nuclear weapons proliferation. Finally, current and future global demand for LWR technology might be below the minimum needed to sustain the current global LWR industry. (author)

Towards intrinsically safe light-water reactors

International Nuclear Information System (INIS)

Hannerz, K.

1983-02-01

The reactor-safety issue is one of the principal problems threatening the future of the nuclear option, at least in participatory democracies. It has contributed to widespread public distrust and is the direct cause of the escalation in design complexity and quality assurance requirements that are rapidly eroding the competitive advantage of nuclear power. Redesign of the light-water reactor can eliminate those features that leave it open to public distrust and obstructive intervention. This redesign appears feasible within the realm of proven technology in those fields (fuels, materials, water chemistry, waste technology, etc.) in which extended operating experience is essential for confidence in system performance. A pressurized water reactor outline design developed to achieve the above goal is presented. The key feature is the design of the primary system extracting heat from the core so that the latter is protected from damage caused by any credible system failure or any destructive intervention from the outside by either violent means (up to and including nonnuclear warfare) or by mistaken or malicious use of the plant control systems. Such a design objective can be achieved by placing the entire primary circulation system in a large pressurized pool of cold water with a high boric acid content. Enough water is provided in the pool to allow core-decay-heat removal by evaporation for at least one week following any incident with no cooling systems operating. Subsequently it is assumed that a supply of further water (a few cubic meters per hour) from the outside can be arranged, even without the presence of the plant operating personnel

Erosion and break-up of light-gas layers by a horizontal jet in a multi-vessel, large-scale containment test system

Energy Technology Data Exchange (ETDEWEB)

Zboray, Robert, E-mail: robert.zboray@psi.ch; Mignot, Guillaume; Kapulla, Ralf; Paladino, Domenico

2015-09-15

The distribution and eventual stratification of hydrogen released during a hypothetical severe accident and the stability of the stratification formed in the early phase of the transient is of particular safety concern in Light Water Reactors (LWRs). The large-scale containment test facility PANDA (PSI, Switzerland) has been used to perform a series of four tests examining the erosion and break-up of stratified light-gas layers in the frame of the OECD SETH-2 project. The ultimate goal of the test program is to set-up an experimental data base of high-quality and high-density data that can challenge and validate 3D containment codes like e.g. GOTHIC, GASFLOW or MARS and validate the applicability of CFD codes like FLUENT or CFX for LWR containment problems. The test series discussed here focuses on the erosion of a stratified, helium-rich layer by horizontal steam injection at different locations below the layer. An approach with step-wise increasing complexity has been chosen to examine this problem allowing control over the rate of pressure increase and the occurrence of condensation. The step-wise approach enables a thorough understanding of the influence of different phenomena like position of steam injection, diffusion, pressurization and condensation on the behavior and erosion of the stratified layer.

Ultraviolet light in the use of water disinfection

International Nuclear Information System (INIS)

Dabbagh, R.

1999-01-01

Ultraviolet light is an effective method in the use of water disinfection for swimming pools, potable water and industry required water. For many reasons Ultraviolet light and Ultraviolet compounded with chlorine (Ultraviolet/chlorine) has been brought to attention ed in resent years. In this research, a swimming pool water disinfection was carried out by means of a system with the use of a reactor which was made of stainless steel (SS-304) and with many another standards required. Operation of system was carried out at first in the pilot plant and then installation in essential water treatment integrated. Inactivation of pollution index, E. Coli or Total coliform and Pseudomonas aeroginosa studies with 6000,16000 and 30000 μW.s/cm 2 Ultraviolet dose and then in presence of 0.3,0.6,0.9 and 1.2 mg/1 free chlorine (Ultraviolet/chlorine). In swimming pools minimum free chlorine residual usually is 1.5 mg/1. Optimum Ultraviolet dose was 16000 μW.s/cm 2 attention to 50 percent Ultraviolet absorption ca sued to TSS,TDS and turbidity. In the Ultraviolet/chlorine system suitable rate was 16000μW.s/cm 2 Ultraviolet dose/0.6 mg/1 chlorine in the 2.4 * 10 5 CFU/100 ml for Total coliform and 3600 CFU/100 ml for Pseudomonas aeroginosa. Most probable number (MPN) estimated multiple tube fermentation technique. In this way the flow rate for system indicated about 240 cm 3 /s or 0.9 m 3 /h. The samples polluted for secondary pollution with 54000 CFU/100 ml for E. Coli and 1800 CFU/100ml Pseudomonas aeroginosa. The number of microbes decreased to zero duration after 45 minutes contact time in presence of free chlorine residual in samples. In practical conditions which that disinfectant system was installed in essential water treatment circuit under 1.4 atm hydraulic pressure no growth was seen for pollution index in disinfected water with Ultraviolet in microbial density about 840 CFU/100 ml for Total coliform and 12 CFU/100 ml for pseudomonas aeroginosa. Attention to lower

Ultraviolet light in the use of water disinfection

International Nuclear Information System (INIS)

Dabbagh, R.

1999-01-01

Ultraviolet light is an effective method in the use of water disinfection for swimming pools, potable water and industry required water. For many reasons UV light and UV compounded with chlorine (UV/chlorine) has been brought to attention in resent years. In this research, a swimming pool water disinfection was carried out by means of a system with the use of a reactor which was made of stainless steel (SS-304) and with many another standards required. Operation of system was carried out at first in the pilot plant and then installation in essential water treatment integrated. Inactivation of pollution index, E. Coli or Total coliform and Pseudomonas aeroginosa studied with 6000,16000 and 30000 μW.s/cm 2 UV dose and then in presence of 0.3,0.6,0.9 and 1.2 mg/1 free chlorine (UV/chlorine). In swimming pools minimum free chlorine residual usually is 1.5 mg/1. Optimum UV dose was 16000 μW.s/cm 2 attention to 50 percent UV absorption caused to TSS,TDS and turbidity. In the UV/chlorine system suitable rate was 16000μW.s/cm 2 UV dose /0.6 mg/1 chlorine in the 2.4 * 10 5 CFU/100 ml for Total coliform and 3600CFU/100 ml for Pseudomonas aeroginosa. Most probable number(MPN) estimated multiple tube fermentation technique. In this way the flow rate for system indicated about 240 cm 3 /s or 0.9 m 3 /h. The samples polluted for secondary pollution with 54000 CFU/100 ml for E.Coli and 1800 CFU/100ml Pseudomonas aeroginosa. The number of microbes decreased to zero duration after 45 minutes contact time in presence of free chlorine residual in samples. In practical conditions which that disinfectant system was installed in essential water treatment circuit under 1.4 atm hydraulic pressure no growth was seen for pollution index in disinfected water with UV in microbial density about 840 CFU/100 ml for Total coliform and 12CFU/100 ml for Pseudomonas aeroginosa. Attention to lower turbidity, TSS and TDS in tap water, higher flow rate about 560 cm 3 /s or 2 m 3 /h acessesed

77 FR 15812 - Initial Test Program of Condensate and Feedwater Systems for Light-Water Reactors

Science.gov (United States)

2012-03-16

... Systems for Light-Water Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Draft regulatory guide... Feedwater Systems for Light- Water Reactors.'' DG-1265 is proposed revision 2 of Regulatory Guide 1.68.1... Plants,'' dated January 1977. This regulatory guide is being revised to: (1) expand the scope of the...

Stationary PWR-calculations by means of LWRSIM at the NEACRP 3D-LWRCT benchmark

International Nuclear Information System (INIS)

Van de Wetering, T.F.H.

1993-01-01

Within the framework of participation in an international benchmark, calculations were executed by means of an adjusted version of the computer code Light Water Reactor SIMulation (LWRSIM) for three-dimensional reactor core calculations of pressurized water reactors. The 3-D LWR Core Transient Benchmark was set up aimed at the comparison of 3-D computer codes for transient calculations in LWRs. Participation in the benchmark provided more insight in the accuracy of the code when applied for other pressurized water reactors than applied for the nuclear power plant Borssele in the Netherlands, for which the code has been developed and used originally

Advanced Reactor Development in the United States

Energy Technology Data Exchange (ETDEWEB)

Giessing, D. F.; Griffith, J. D.; McGoff, D. J.; Rosen, Sol [U. S. Department of Energy, Texas (United States)

1990-04-15

In the United States, three technologies are employed for the new generation of advanced reactors. These technologies are Advanced Light Water Reactors (A LWRs) for the 1990s and beyond, the Modular High Temperature Gas Reactor (M HTGR) for commercial use after the turn of the century, and Liquid Metal Reactors (LWRs) to provide energy production and to convert reactor fission waste to a more manageable waste product. Each technology contributes to the energy solution. Light Water Reactors For The 1990s And Beyond--The U. S. Program The economic and national security of the United States requires a diversified energy supply base built primarily upon adequate, domestic resources that are relatively free from international pressures. Nuclear energy is a vital component of this supply and is essential to meet current and future national energy demands. It is a safe, economically continues to contribute to national energy stability, and strength. The Light Water Reactor (LWR) has been a major and successful contributor to the electrical generating needs of many nations throughout the world. It is being counted upon in the United States as a key to revitalizing nuclear energy option in the 1990s. In recent years, DOE joined with the industry to ensure the availability and future viability of the LWR option. This national program has the participation of the Nation's utility industry, the Electric Power Research Institute (EPRI), and several of the major reactor manufacturers and architect-engineers. Separate but coordinated parts of this program are managed by EPRI and DOE.

Zinc oxide nanorod mediated visible light photoinactivation of model microbes in water

Energy Technology Data Exchange (ETDEWEB)

Sapkota, Ajaya; Anceno, Alfredo J; Dutta, Joydeep [Center of Excellence in Nanotechnology, Asian Institute of Technology, Klong Luang, Pathumthani 12120 (Thailand); Baruah, Sunandan; Shipin, Oleg V, E-mail: alfredo.anceno@cemagref.fr, E-mail: joy@ait.ac.th [Environmental Engineering and Management, Asian Institute of Technology, Klong Luang, Pathumthani 12120 (Thailand)

2011-05-27

The inactivation of model microbes in aqueous matrix by visible light photocatalysis as mediated by ZnO nanorods was investigated. ZnO nanorods were grown on glass substrate following a hydrothermal route and employed in the inactivation of gram-negative Escherichia coli and gram-positive Bacillus subtilis in MilliQ water. The concentration of Zn{sup 2+} ions in the aqueous matrix, bacterial cell membrane damage, and DNA degradation at post-exposure were also studied. The inactivation efficiencies for both organisms under light conditions were about two times higher than under dark conditions across the cell concentrations assayed. Anomalies in supernatant Zn{sup 2+} concentration were observed under both conditions as compared to control treatments, while cell membrane damage and DNA degradation were observed only under light conditions. Inactivation under dark conditions was hence attributed to the bactericidal effect of Zn{sup 2+} ions, while inactivation under light conditions was due to the combined effects of Zn{sup 2+} ions and photocatalytically mediated electron injection. The reduction of pathogenic bacterial densities by the photocatalytically active ZnO nanorods in the presence of visible light implies potential ex situ application in water decontamination at ambient conditions under sunlight.

Biomimetic Water-Collecting Fabric with Light-Induced Superhydrophilic Bumps.

Science.gov (United States)

Wang, Yuanfeng; Wang, Xiaowen; Lai, Chuilin; Hu, Huawen; Kong, Yeeyee; Fei, Bin; Xin, John H

2016-02-10

To develop an efficient water-collecting surface that integrates both fast water-capturing and easy drainage properties is of high current interest for addressing global water issues. In this work, a superhydrophobic surface was fabricated on cotton fabric via manipulation of both the surface roughness and surface energy. This was followed by a subsequent spray coating of TiO2 nanosol that created light-induced superhydrophilic bumps with a unique raised structure as a result of the interfacial tension of the TiO2 nanosol sprayed on the superhydrophobic fiber surface. These raised TiO2 bumps induce both a wettability gradient and a shape gradient, synergistically accelerating water coalescence and water collection. The in-depth study revealed that the quantity and the distribution of the TiO2 had a significant impact on the final water collection efficiency. This inexpensive and facilely fabricated fabric biomimicks the desert beetle's back and spider silk, which are capable of fog harvesting without additional energy consumption.

Electrochemistry in light water reactors reference electrodes, measurement, corrosion and tribocorrosion issues

CERN Document Server

Bosch, R -W; Celis, Jean-Pierre

2007-01-01

There has long been a need for effective methods of measuring corrosion within light water nuclear reactors. This important volume discusses key issues surrounding the development of high temperature reference electrodes and other electrochemical techniques. The book is divided into three parts with part one reviewing the latest developments in the use of reference electrode technology in both pressurised water and boiling water reactors. Parts two and three cover different types of corrosion and tribocorrosion and ways they can be measured using such techniques as electrochemical impedance spectroscopy. Topics covered across the book include in-pile testing, modelling techniques and the tribocorrosion behaviour of stainless steel under reactor conditions. Electrochemistry in light water reactors is a valuable reference for all those concerned with corrosion problems in this key technology for the power industry. Discusses key issues surrounding the development of high temperature reference eletrodes A valuab...

Investigations on the gas distribution phenomena inside the containment system of LWRs

International Nuclear Information System (INIS)

Manfredini, A.; Oriolo, F.; Villotti, A.

1994-01-01

The importance of mixing and distribution phenomena of hydrogen gas in the reactor safety is emphasised in the advanced reactor concepts, that heavily rely upon the passive cooling systems during a typical severe accident sequence. An advanced methodology for evaluating the temporal and spatial distribution of non condensable gases, including the simulation of buoyancy-driven flows and the effects of the various ESFs activation, in a multi-compartment containment system of a LWR is reviewed. The methodology employs an analogy technique with electrical networks to determine the convection flows among the containment compartments and evaluates, inside a single node, the profile of the vertical concentrations of steam and non condensable gases. The application of the proposed models to simulate the gas distribution phenomena occurring in the HDR E11.2, in the FIPLOC-F2 and in the NUPEC M-7-1 tests demonstrates the importance of these models providing information about local details and spatial distribution. The main results from the post-test analysis performed to simulate the thermal-hydraulic responses of the above mentioned experiments are presented and demonstrate the improvements and the reduction of the error band with respect to the experimental data. This methodology allows to perform a realistic prediction of severe accident sequence inside the containment system of the actual and advanced passive generation of LWRs. (author). 14 refs., 11 figs

Using the Human Systems Simulation Laboratory at Idaho National Laboratory for Safety Focused Research

Energy Technology Data Exchange (ETDEWEB)

Joe, Jeffrey .C; Boring, Ronald L.

2016-07-01

Under the United States (U.S.) Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) program, researchers at Idaho National Laboratory (INL) have been using the Human Systems Simulation Laboratory (HSSL) to conduct critical safety focused Human Factors research and development (R&D) for the nuclear industry. The LWRS program has the overall objective to develop the scientific basis to extend existing nuclear power plant (NPP) operating life beyond the current 60-year licensing period and to ensure their long-term reliability, productivity, safety, and security. One focus area for LWRS is the NPP main control room (MCR), because many of the instrumentation and control (I&C) system technologies installed in the MCR, while highly reliable and safe, are now difficult to replace and are therefore limiting the operating life of the NPP. This paper describes how INL researchers use the HSSL to conduct Human Factors R&D on modernizing or upgrading these I&C systems in a step-wise manner, and how the HSSL has addressed a significant gap in how to upgrade systems and technologies that are built to last, and therefore require careful integration of analog and new advanced digital technologies.

Coatings used in light-water nuclear power plants

International Nuclear Information System (INIS)

Anon.

1981-01-01

The guide is intended to provide a common basis in the selection of test methods which may be required to evaluate and qualify protective coatings (paints) to be used in a light-water nuclear power plant. Standard test methods for the determination of fire resistance, chemical resistance, physical properties, effects of radiation, decontaminability, thermal conductivity, repairability, and for evaluation under accident conditions are included

Loose parts monitoring in light water reactor cooling systems

International Nuclear Information System (INIS)

Santos, A.; Alma, B.J.

1982-01-01

The work related to loose monitoring system for light water reactor, developed at GRS - Munique, are described. The basic problems due to the exact localization and detection of the loose part as well the research activities and development necessary aiming to obtain the best techniques in this field. (E.G.) [pt

Supercritical-pressure, once-through cycle light water cooled reactor concept

International Nuclear Information System (INIS)

Oka, Yoshiaki; Koshizuka, Seiichi

2001-01-01

The purpose of the study is to develop new reactor concepts for the innovation of light water reactors (LWR) and fast reactors. Concept of the once-through coolant cycle, supercritical-pressure light water cooled reactor was developed. Major aspects of reactor design and safety were analysed by the computer codes which were developed by ourselves. It includes core design of thermal and fast reactors, plant system, safety criteria, accident and transient analysis, LOCA, PSA, plant control, start up and stability. High enthalpy rise as supercritical boiler was achieved by evaluating the cladding temperature directly during transients. Fundamental safety principle of the reactor is monitoring coolant flow rate instead of water level of LWR. The reactor system is compact and simple because of high specific enthalpy of supercritical water and the once-through cycle. The major components are similar to those of LWR and supercritical thermal plant. Their temperature are within the experiences in spite of the high outlet coolant temperature. The reactor is compatible with tight fuel lattice fast reactor because of the high head pumps and low coolant flow rate. The power rating of the fast reactor is higher than the that of thermal reactor because of the high power density. (author)

Practical considerations in the use of UV light for drinking water disinfection

International Nuclear Information System (INIS)

Jeyanayagam, S.; Cotton, C.

2002-01-01

Ultraviolet (UV) light was discovered approximately 150 years ago. The first commercial UV lamp was made in the early 1900s soon followed by the manufacture of the quartz sleeve. These technological advances allowed the first application of UV light for water disinfection in 1907 in France. In the mid 1980s, UV disinfection was named as a Best available technology (BAT) for wastewater disinfection in the United States. Fueled by the recent findings that UV disinfection can inactivate key pathogens at cost effective UV doses, the drinking water industry in North America is closely investigating its application in large installations. (author)

Investigation into the Use of Water Based Brake Fluid for Light Loads

Directory of Open Access Journals (Sweden)

W. A. Akpan

2012-12-01

Full Text Available This paper addresses the possibility of using water based fluid as a brake fluid for light loads. Characterization of both standard and water based braked fluids formulated was carried out. The properties of the latter were compared with that of a standard commercial brake fluid. The actual test of the formulated brake fluid was carried out with a Nissan Sunny vehicle model 1.5 within the speed range of 20km/hr to 80km/hr at the permanent campus of University of Uyo and the braking efficiency obtained attest to its suitability for light loads.

MOX recycling-an industrial reality

International Nuclear Information System (INIS)

Shallo, G.D.F.

1996-01-01

Reprocessing and plutonium recycling have now attained industrial maturity in France and Europe. Specifically, mixed-oxide (MOX) fuel is fabricated and used in light water reactors (LWRs) in satisfactory operating conditions. The utilities and the fuel cycle industry experience no technical difficulties, and European recycling programs are growing steadily, from 18 reactors in operation today up to 50 expected around the year 2000, putting the system reprocessing-recycling in coherence: 25 t of plutonium will then be used each year to produce the electricity equivalence of 25 millions tons of oil. Plutonium recycling in MOX fuel in current LWRs proves to be technically safe and economically competitive and meets natural resource savings and environmental protection objectives. And recycling responds properly to the nonproliferation concerns. Such an industrial experience gives a unique reference for weapons plutonium disposition through MOX use in reactors

Analysis on Radioactive Waste Transmutation in Light Water cooled Hyb-WT

International Nuclear Information System (INIS)

Hong, Seonghee; Kim, Myung Hyun

2014-01-01

A feasibility of realization is much higher in FFHR compared with pure fusion. A combination of plasma fusion source for neutrons with a subcritical reactor at the blanket side has much higher capability in transmutation of waste as well as reactor safety compared with fission reactor options. Fusion-Fission Hybrid Reactor (FFHR) uses various coolants depending on the purpose. It is important that coolant being used should be suitable to reactor purpose, because reactor performance and the design constraints may change depending on the coolant. There are basically two major groups of coolants for FFHR. One group of coolant does not contain Li. They are Na, Pb-Bi, H 2 O and D 2 O. The other group contains Li for tritium breeding. They are Li, LiPb, LiSN, FLIBE and FLiNaBe. Currently, the issue in FFHR is its implication for radioactive waste transmutation (FFHR for WT). Because radioactive wastes of spent nuclear fuel (SNF) are transmuted using fusion neutron source. Therefore a suitable coolant should be used for effective waste transmutation. . In FFHR for WT, LiPb coolant is being used mainly because of tritium production in Li and high neutron economic through reaction in Pb. However different coolants use such as Na, Pb-Bi are used in fast reactors and accelerator driven systems (ADS) having same purpose. In this study, radioactive waste transmutation performance of various coolants mentioned above will be compared and analyzed. Through this study, the coolants are judged primarily for their support to waste transmutation disregarding their limitation to reactor design and tritium breeding capability. First, performance of the light water coolant regarding radioactive waste transmutation was analyzed among various coolants mentioned above. In this paper, performance of radioactive waste transmutation can be known depending on different volume fractions (54.53, 60.27, 97.94vol.%) of the light water. Light water dose required fusion power lower than LiPb due to

Dual-purpose light water reactor supplying heat for desalination

International Nuclear Information System (INIS)

Waplington, G.; Fichtner, H.

1978-01-01

The technical as well as the economic aspects of using a large commercial light water reactor for the production of both electricity and potable water have been examined. For the basis of the study, the multistage flash distillation process was selected, in conjunction with a reactor rated at not less than 2100 MW (thermal). Combined use of a condensing and a back-pressure turbine (the latter matched to distillation plant steam requirements) represents a convenient method for supplying process heat. Overall costs can be fairly allocated to the two products using the ''power credit'' method. A sample economic evaluation indicates highly favorable water costs as compared with more conventional distillation schemes based on fossil fuel

Overview of environmental materials degradation in light-water reactors

International Nuclear Information System (INIS)

Shaaban, H.I.; Wu, P.

1986-08-01

This report provides a brief overview of analyses and conclusions reported in published literature regarding environmentally induced degradation of materials in operating light-water reactors. It is intended to provide a synopsis of subjects of concern rather than to address a licensing basis for any newly discovered problems related to reactor materials

Development trends in light water reactors

International Nuclear Information System (INIS)

Fogelstroem, L.; Simon, M.

1988-01-01

The present market for new nuclear power plants is weak, but is expected to pick up again, which is why great efforts are being made to further develop the light water reactor line for future applications. There is both a potential and a need for further improvement, for instance with respect to even higher cost efficiency, a simplified operating permit procedure, shorter construction periods, and increased operational flexibility to meet rising demands in load following behavior and in better cycle data of fuel elements. However, also public acceptance must not be forgotten when deciding about the line to be followed in the development of LWR technology. (orig.) [de

Is light water reactor technology sustainable?

Energy Technology Data Exchange (ETDEWEB)

Rothwell, G. [Stanford Univ., Dept. of Economics, CA (United States); Van der Zwaan, B. [Vrije Univ., Amsterdam, Inst. for Environmental Studies (Netherlands)

2001-07-01

This paper proposes criteria for determining ''intermediate sustainability'' over a 500-year horizon. We apply these criteria to Light Water Reactor (LWR) technology and the LWR industry. We conclude that LWR technology does not violate intermediate sustainability criteria for (1) environmental externalities, (2) worker and public health and safety, or (3) accidental radioactive release. However, it does not meet criteria to (1) efficiently use depleted uranium and (2) avoid uranium enrichment technologies that can lead to nuclear weapons proliferation. Finally, current and future global demand for LWR technology might be below the minimum needed to sustain the current global LWR industry. (author)

Light pollution offshore: Zenithal sky glow measurements in the mediterranean coastal waters

Science.gov (United States)

Ges, Xavier; Bará, Salvador; García-Gil, Manuel; Zamorano, Jaime; Ribas, Salvador J.; Masana, Eduard

2018-05-01

Light pollution is a worldwide phenomenon whose consequences for the natural environment and the human health are being intensively studied nowadays. Most published studies address issues related to light pollution inland. Coastal waters, however, are spaces of high environmental interest, due to their biodiversity richness and their economical significance. The elevated population density in coastal regions is accompanied by correspondingly large emissions of artificial light at night, whose role as an environmental stressor is increasingly being recognized. Characterizing the light pollution levels in coastal waters is a necessary step for protecting these areas. At the same time, the marine surface environment provides a stage free from obstacles for measuring the dependence of the skyglow on the distance to the light polluting sources, and validating (or rejecting) atmospheric light propagation models. In this work we present a proof-of-concept of a gimbal measurement system that can be used for zenithal skyglow measurements on board both small boats and large vessels under actual navigation conditions. We report the results obtained in the summer of 2016 along two measurement routes in the Mediterranean waters offshore Barcelona, travelling 9 and 31.7 km away from the coast. The atmospheric conditions in both routes were different from the ones assumed for the calculation of recently published models of the anthropogenic sky brightness. They were closer in the first route, whose results approach better the theoretical predictions. The results obtained in the second route, conducted under a clearer atmosphere, showed systematic differences that can be traced back to two expected phenomena, which are a consequence of the smaller aerosol content: the reduction of the anthropogenic sky glow at short distances from the sources, and the slower decay rate of brightness with distance, which gives rise to a relative excess of brightness at large distances from the

10-fold enhancement in light-driven water splitting using niobium oxynitride microcone array films

KAUST Repository

Shaheen, Basamat

2016-03-26

We demonstrate, for the first time, the synthesis of highly ordered niobium oxynitride microcones as an attractive class of materials for visible-light-driven water splitting. As revealed by the ultraviolet photoelectron spectroscopy (UPS), photoelectrochemical and transient photocurrent measurements, the microcones showed enhanced performance (~1000% compared to mesoporous niobium oxide) as photoanodes for water splitting with remarkable stability and visible light activity. © 2016 Elsevier B.V. All rights reserved.

Aging assessment and mitigation for major LWR [light water reactor] components

International Nuclear Information System (INIS)

Shah, Y.N.; Ware, A.G.; Conley, D.A.; MacDonald, P.E.; Burns, J.J. Jr.

1989-01-01

This paper summarizes some of the results of the Aging Assessment and Mitigation Project sponsored by the US Nuclear Regulatory Commission (USNRC), Office of Nuclear Regulatory Research. The objective of the project is to develop an understanding of the aging degradation of the major light water reactor (LWR) structures and components and to develop methods for predicting the useful life of these components so that the impact of aging on the safe operation of nuclear power plants can be evaluated and addressed. The research effort consists of integrating, evaluating, and updating the available aging-related information. This paper discusses current accomplishments and summarizes the significant degradation processes active in two major components: pressurized water reactor pressurizer surge and spray lines and nozzles, and light water reactor primary coolant pumps. This paper also evaluates the effectiveness of the current inservice inspection programs and presents conclusions and recommendations related to aging of these two major components. 37 refs., 7 figs., 3 tabs

Remote monitoring technical review for light water reactors (Phase 1)

Energy Technology Data Exchange (ETDEWEB)

Park, Seung Sik; Yoon, Wan Ki; Na, Won Woo; Kwack, Eun Ho [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-10-01

The IAEA has been conducting a field trial of a Remote Monitoring System (RMS) at the spent fuel storage, Younggwang 3 nuclear power plant. The system installation plan was initiated after the agreement in the 7th ROK-IAEA safeguards Implementation Review Meeting that was held in Soul, 1998. It describes that IAEA and Korea proceed RM tasks Implementation of RMS at LWRs in the ROK for field trials. The project of RMS is conducting through 3 stages with timing. RMS has been installed for the Phase I of field trial, one of two stages at Younggwang Unit 3 in October 1998. The RMS consists of video systems and a seal at the spent fuel pond area. This report provides a description of the monitoring system and its functions focusing on several technical points of the installation and its 6 month operation at Younggwang Unit 3. Subjects are selected and analyzed in the three chapters, IAEA safeguards policy on Remote Monitoring, the technology, and field test experiences. 8 refs., 12 figs., 12 tabs. (Author)

Analysis of the surveillance test data on irradiation embrittlement of the reactor pressure vessel steels in LWRs

International Nuclear Information System (INIS)

Lee, Gyoeng Geun; Kwon, Jun Hyun

2010-11-01

The surveillance test data in Korean LWRs were analyzed from a viewpoint of materials science. TTS change with the neutron irradiation were compared to the model values of the RG1.99/2 and NUREG/CR-6551. The model values of TTS were higher than the actual values of TTS. It was impossible to find a relationship between TTS and neutron fluence in weld data. The correlation of the increase in YS (yield strength) and TTS with neutron irradiation was also investigated. Like the result of TTS change, the YS/TTS showed the correlations in plate/forgings metals, however no correlation in weld metals. The data were similar to Odette's result about US surveillance tests. From the empirical relationships, the TTS curve change could be predicted using the CVN test result of the unirradiated specimen and the change in YS with neutron irradiation of the specimen

Estimation of diffuse attenuation of ultraviolet light in optically shallow Florida Keys waters from MODIS measurements

Science.gov (United States)

Diffuse attenuation of solar light (Kd, m−1) determines the percentage of light penetrating the water column and available for benthic organisms. Therefore, Kd can be used as an index of water quality for coastal ecosystems that are dependent on photosynthesis, such as the coral ...

RETRAN sensitivity studies of light water reactor transients. Final report

International Nuclear Information System (INIS)

Burrell, N.S.; Gose, G.C.; Harrison, J.F.; Sawtelle, G.R.

1977-06-01

This report presents the results of sensitivity studies performed using the RETRAN/RELAP4 transient analysis code to identify critical parameters and models which influence light water reactor transient predictions. Various plant transients for both boiling water reactors and pressurized water reactors are examined. These studies represent the first detailed evaluation of the RETRAN/RELAP4 transient code capability in predicting a variety of plant transient responses. The wide range of transients analyzed in conjunction with the parameter and modeling studies performed identify several sensitive areas as well as areas requiring future study and model development

Hydrogen evolution from water using solid carbon and light energy

Energy Technology Data Exchange (ETDEWEB)

Kawai, T; Sakata, T

1979-11-15

Hydrogen is produced from water vapour and solid carbon when mixed powders of TiO2, RuO2 and active carbon exposed to water vapor at room temperature, or up to 80 C, are illuminated. At 80 C, the rate of CO and COat2 formation increased. Therefore solar energy would be useful here as a combination of light energy and heat energy. Oxygen produced on the surface of the photocatalyst has a strong oxidising effect on the carbon. It is suggested that this process could be used for coal gasification and hydrogen production from water, accompanied by storage of solar energy.

Radioactivity, radiation protection and monitoring during dismantling of light-water reactors

International Nuclear Information System (INIS)

Hummel, L.; Zech, J.B.

2005-01-01

Based on the radioactivity inventory in the systems and components of light-water reactors observed during operation, the impact of actions during plant emptying after the conclusion of power operation and possible subsequent long-term safe enclosure concerning the composition of the nuclide inventory of the plant to be dismantled will be described. Derived from this will be the effects on radioactivity monitoring in the plant, physical radiation protection monitoring, and the measured characterization of the residual materials resulting from the dismantling. The impact of long-term interim storage will also be addressed in the discussion. The talk should provide an overview of the interrelationships between source terms, decay times and the radioactivity monitoring requirements of the various dismantling concepts for commercial light-water reactors. (orig.)

NEPTUNE: a modular system for light-water reactor calculation

International Nuclear Information System (INIS)

Bouchard, J.; Kanevoky, A.; Reuss, P.

1975-01-01

A complete modular system of light water reactor calculations has been designed. It includes basic nuclear data processing, the APOLLO phase: transport calculations for cells, multicells, fuel assemblies or reactors, the NEPTUNE phase: reactor calculations. A fuel management module, devoted to the automatic determination of the best shuffling strategy is included in NEPTUNE [fr

Removal and recovery of tritium from light and heavy water

International Nuclear Information System (INIS)

Butler, J.P.; Hammerli, M.

1979-01-01

A method and apparatus for removing tritium from light water are described, comprising contacting tritiated feed water in a catalyst column in countercurrent flow with hydrogen gas originating from an electrolysis cell so as to enrich this feed water with tritium from the electrolytic hydrogen gas and passing the tritium enriched water to an electrolysis cell wherein the electrolytic hydrogen gas is generated and then fed upwards through the catalyst column or recovered as product. The tritium content of the hydrogen gas leaving the top of the enricher catalyst column is further reduced in a stripper column containing catalyst which transfers the tritium to a countercurrent flow of liquid water. Anodic oxygen and water vapour from the anode compartment may be fed to a drier and condensed electrolyte recycled with a slip stream or recovered as a further tritium product stream. A similar method involving heavy water is also described. (author)

Validation of the Nuclear Design Method for MOX Fuel Loaded LWR Cores

International Nuclear Information System (INIS)

Saji, E.; Inoue, Y.; Mori, M.; Ushio, T.

2001-01-01

The actual batch loading of mixed-oxide (MOX) fuel in light water reactors (LWRs) is now ready to start in Japan. One of the efforts that have been devoted to realizing this batch loading has been validation of the nuclear design methods calculating the MOX-fuel-loaded LWR core characteristics. This paper summarizes the validation work for the applicability of the CASMO-4/SIMULATE-3 in-core fuel management code system to MOX-fuel-loaded LWR cores. This code system is widely used by a number of electric power companies for the core management of their commercial LWRs. The validation work was performed for both boiling water reactor (BWR) and pressurized water reactor (PWR) applications. Each validation consists of two parts: analyses of critical experiments and core tracking calculations of operating plants. For the critical experiments, we have chosen a series of experiments known as the VENUS International Program (VIP), which was performed at the SCK/CEN MOL laboratory in Belgium. VIP consists of both BWR and PWR fuel assembly configurations. As for the core tracking calculations, the operating data of MOX-fuel-loaded BWR and PWR cores in Europe have been utilized

Design features to facilitate IAEA safeguards at light water reactors

International Nuclear Information System (INIS)

Pasternak, T.; Glancy, J.; Goldman, L.; Swartz, J.

1981-01-01

Several studies have been performed recently to identify and analyze light water reactor (LWR) features that, if incorporated into the facility design, would facilitate the implementation of International Atomic Energy Agency (IAEA) safeguards. This paper presents results and conclusions of these studies. 2 refs

The significance of water hammer events to public dose from reactor accidents: A probabilistic assessment

International Nuclear Information System (INIS)

Amico, P.J.; Ferrell, W.L.; Rubin, M.P.

1984-01-01

A probabilistic assessment was made of the effects on public dose of water hammer events in LWRs. The analysis utilized actual historical water hammer data to determine if the water hammer events contributed either to system failure rates or initiating event frequencies. Representative PRAs were used to see if changes in initiating events and/or system failures caused by water hammer resulted in new values for the dominant sequences in the PRAs. New core melt frequencies were determined and carried out to the subsequent increase in public dose. It is concluded that water hammer is not a significant problem with respect to risk to the public for either BWRs or PWRs. (orig./HP)

Ultraviolet light-emitting diodes in water disinfection.

Science.gov (United States)

Vilhunen, Sari; Särkkä, Heikki; Sillanpää, Mika

2009-06-01

The novel system of ultraviolet light-emitting diodes (UV LEDs) was studied in water disinfection. Conventional UV lamps, like mercury vapor lamp, consume much energy and are considered to be problem waste after use. UV LEDs are energy efficient and free of toxicants. This study showed the suitability of LEDs in disinfection and provided information of the effect of two emitted wavelengths and different test mediums to Escherichia coli destruction. Common laboratory strain of E. coli (K12) was used and the effects of two emitted wavelengths (269 and 276 nm) were investigated with two photolytic batch reactors both including ten LEDs. The effects of test medium were examined with ultrapure water, nutrient and water, and nutrient and water with humic acids. Efficiency of reactors was almost the same even though the one emitting higher wavelength had doubled optical power compared to the other. Therefore, the effect of wavelength was evident and the radiation emitted at 269 nm was more powerful. Also, the impact of background was studied and noticed to have only slight deteriorating effect. In the 5-min experiment, the bacterial reduction of three to four log colony-forming units (CFU) per cubic centimeter was achieved, in all cases. When turbidity of the test medium was greater, part of the UV radiation was spent on the absorption and reactions with extra substances on liquid. Humic acids can also coat the bacteria reducing the sensitivity of the cells to UV light. The lower wavelength was distinctly more efficient when the optical power is considered, even though the difference of wavelengths was small. The reason presumably is the greater absorption of DNA causing more efficient bacterial breakage. UV LEDs were efficient in E. coli destruction, even if LEDs were considered to have rather low optical power. The effect of wavelengths was noticeable but the test medium did not have much impact. This study found UV LEDs to be an optimal method for bacterial

Radiation Protection at Light Water Reactors

CERN Document Server

Prince, Robert

2012-01-01

This book is aimed at Health Physicists wishing to gain a better understanding of the principles and practices associated with a light water reactor (LWR) radiation protection program. The role of key program elements is presented in sufficient detail to assist practicing radiation protection professionals in improving and strengthening their current program. Details related to daily operation and discipline areas vital to maintaining an effective LWR radiation protection program are presented. Programmatic areas and functions important in preventing, responding to, and minimizing radiological incidents and the importance of performing effective incident evaluations and investigations are described. Elements that are integral in ensuring continuous program improvements are emphasized throughout the text.

Aerosol behavior and light water reactor source terms

International Nuclear Information System (INIS)

Abbey, F.; Schikarski, W.O.

1988-01-01

The major developments in nuclear aerosol modeling following the accident to pressurized water reactor Unit 2 at Three Mile Island are briefly reviewed and the state of the art summarized. The importance and implications of these developments for severe accident source terms for light water reactors are then discussed in general terms. The treatment is not aimed at identifying specific source term values but is intended rather to illustrate trends, to assess the adequacy of the understanding of major aspects of aerosol behavior for source term prediction, and demonstrate in qualitative terms the effect of various aspects of reactor design. Areas where improved understanding of aerosol behavior might lead to further reductions in current source terms predictions are also considered

Investigation of valve failure problems in LWR power plants

International Nuclear Information System (INIS)

1980-04-01

An analysis of component failures from information in the computerized Nuclear Safety Information Center (NSIC) data bank shows that for both PWR and BWR plants the component category most responsible for approximately 19.3% of light water reactor (LWR) power plant shutdowns. This investigation by Burns and Roe, Inc. shows that the greatest cause of shutdowns in LWRs due to valve failures is leakage from valve stem packing. Both BWR plants and PWR plants have stem leakage problems

Current and anticipated uses of thermal hydraulic codes at the Japan Atomic Energy Research Institute

International Nuclear Information System (INIS)

Akimoto, Hajime; Kukita; Ohnuki, Akira

1997-01-01

The Japan Atomic Energy Research Institute (JAERI) is conducting several research programs related to thermal-hydraulic and neutronic behavior of light water reactors (LWRs). These include LWR safety research projects, which are conducted in accordance with the Nuclear Safety Commission's research plan, and reactor engineering projects for the development of innovative reactor designs or core/fuel designs. Thermal-hydraulic and neutronic codes are used for various purposes including experimental analysis, nuclear power plant (NPP) safety analysis, and design assessment

Generic safety issues for nuclear power plants with light water reactors and measures taken for their resolution

International Nuclear Information System (INIS)

1998-09-01

The IAEA Conference on 'The Safety of Nuclear Power: Strategy for the Future' in 1991 was a milestone in nuclear safety. Two of the important items addressed by this conference were ensuring and enhancing safety of operating plants and treatment of nuclear power plants built to earlier safety standards. A number of publications related to these two items issued subsequent to this conference were: A Common Basis for Judging the Safety of Nuclear Power Plants Built to Earlier Standards, INSAG-9 (1995), the IAEA Safety Guide 50-SG-O12, periodic Safety Review of Operational Nuclear Power Plants (1994) and an IAEA publication on the Safety Evaluation of Operating Nuclear Power Plants Built to Earlier Standards - A Common Basis for Judgement (1997). Some of the findings of the 1991 Conference have not yet been fully addressed. An IAEA Symposium on reviewing the Safety of Existing Nuclear Power Plants in 1996 showed that there is an urgent need for operating organizations and national authorities to review operating nuclear power plants which do not meet the high safety levels of the vast majority of plants and to undertake improvements with assistance from the international community if required. Safety reviews of operating nuclear power plants take on added importance in the context of the Convention on Nuclear safety and its implementation. The purpose of this TECDOC compilation based on broad international experience, is to assist the Member States in the reassessment of operating plants by providing a list of generic safety issues identified in nuclear power plants together with measures taken to resolve these issues. These safety issues are generic in nature with regard to light water reactors and the measures for their resolution are for use as a reference for the safety reassessment of operating plants. The TECDOC covers issues thought to be significant to Member States based on consensus process. It provides an introduction to the use of generic safety issues for

CEA studies on advanced nuclear fuel claddings for enhanced accident tolerant LWRs fuel (LOCA and beyond LOCA conditions)

International Nuclear Information System (INIS)

Brachet, J.C.; Lorrette, C.; Michaux, A.; Sauder, C.; Idarraga-Trujillo, I.; Le Saux, M.; Le Flem, M.; Schuster, F.; Billard, A.; Monsifrot, E.; Torres, E.; Rebillat, F.; Bischoff, J.; Ambard, A.

2015-01-01

This paper gives an overview of CEA studies on advanced nuclear fuel claddings for enhanced Accident Tolerant LWR Fuel in collaboration with industrial partners AREVA and EDF. Two potential solutions were investigated: chromium coated zirconium based claddings and SiC/SiC composite claddings with a metallic liner. Concerning the first solution, the optimization of chromium coatings on Zircaloy-4 substrate has been performed. Thus, it has been demonstrated that, due in particular to their slower oxidation rate, a significant additional 'grace period( can be obtained on high temperature oxidized coated claddings in comparison to the conventional uncoated ones, regarding their residual PQ (Post-Quench) ductility and their ability to survive to the final water quenching in LOCA and, to some extent, beyond LOCA conditions. Concerning the second solution, the innovative 'sandwich' SiC/SiC cladding concept is introduced. Initially designed for the next generation of nuclear reactors, it can be adapted to obtain high safety performance for LWRs in LOCA conditions. The key findings of this work highlight the low sensitivity of SiC/SiC composites under the explored steam oxidation conditions. No signification degradation of the mechanical properties of CVI-HNI SiC/SiC specimen is particularly acknowledged for relatively long duration (beyond 100 h at 1200 Celsius degrees). Despite these very positive preliminary results, significant studies and developments are still necessary to close the technology gap. Qualification for nuclear application requires substantial irradiation testing, additional characterization and the definition of design rules applicable to such a structure. The use of a SiC-based fuel cladding shows promise for the highest temperature accident conditions but remains a long term perspective

The manufacture of plutonium fuels for light water reactors

International Nuclear Information System (INIS)

Lebastard, G.

1985-01-01

This paper describes the agreement concluded between COGEMA and BELGONUCLEAIRE, reflected in the creation of the COMMOX group which has been made reponsible for promoting and marketing plutonium fuel rods for light water reactors. One then analyses the main aspects of manufacturing this type of fuel and the resources deployed. Finally one indicates the sales prospects scheduled to meet requirements (MELOX plant) [fr

Analysis of thermal fatigue events in light water reactors

Energy Technology Data Exchange (ETDEWEB)

Okuda, Yasunori [Institute of Nuclear Safety System Inc., Seika, Kyoto (Japan)

2000-09-01

Thermal fatigue events, which may cause shutdown of nuclear power stations by wall-through-crack of pipes of RCRB (Reactor Coolant Pressure Boundary), are reported by licensees in foreign countries as well as in Japan. In this paper, thermal fatigue events reported in anomalies reports of light water reactors inside and outside of Japan are investigated. As a result, it is clarified that the thermal fatigue events can be classified in seven patterns by their characteristics, and the trend of the occurrence of the events in PWRs (Pressurized Water Reactors) has stronger co-relation to operation hours than that in BWRs (Boiling Water Reactors). Also, it is concluded that precise identification of locations where thermal fatigue occurs and its monitoring are important to prevent the thermal fatigue events by aging or miss modification. (author)

Fuel production for LWRs - MOX fuel aspects

International Nuclear Information System (INIS)

Deramaix, P.

2005-01-01

Plutonium recycling in Light Water Reactors is today an industrial reality. It is recycled in the form of (U, Pu)O 2 fuel pellets (MOX), fabricated to a large extent according to UO 2 technology and pellet design. The similarity of physical, chemical, and neutron properties of both fuels also allows MOX fuel to be burnt in nuclear plants originally designed to burn UO 2 . The industrial processes presently in use or planned are all based on a mechanical blending of UO 2 and PuO 2 powders. To obtain finely dispersed plutonium and to prevent high local concentration of plutonium, the feed materials are micronised. In the BNFL process, the whole (UO 2 , PuO 2 ) blend is micronised by attrition milling. According to the MIMAS process, developed by BELGONUCLEAIRE, a primary blend made of UO 2 containing about 30% PuO 2 is micronised in a ball mill, afterwards this primary blend is mechanically diluted in UO 2 to obtain the specified Pu content. After mixing, the (U, Pu)O 2 powder is pressed and the pellets are sintered. The sintering cover gas contains moisture and 5 v/o H 2 . Moisture increases the sintering process and the U-Pu interdiffusion. After sintering and grinding, the pellets are submitted to severe controls to verify conformity with customer specifications (fissile content, Pu distribution, surface condition, chemical purity, density, microstructure). (author)

Investigation into the use of water based brake fluid for light loads ...

African Journals Online (AJOL)

The actual test of the formulated brake fluid was carried out with a Nissan Sunny vehicle model 1.5 within the speed range of 20km/hr to 80km/hr at the permanent campus· of University of Uyo and the· braking effiqiency obtained at test to its suitability for light loads. Keywords·: Water-based, Brake fluid properties, Light loads ...

Nuclear reactors: physics and materials

Energy Technology Data Exchange (ETDEWEB)

Yadigaroglu, G

2005-07-01

In the form of a tutorial addressed to non-specialists, the article provides an introduction to nuclear reactor technology and more specifically to Light Water Reactors (LWR); it also shows where materials and chemistry problems are encountered in reactor technology. The basics of reactor physics are reviewed, as well as the various strategies in reactor design and the corresponding choices of materials (fuel, coolant, structural materials, etc.). A brief description of the various types of commercial power reactors follows. The design of LWRs is discussed in greater detail; the properties of light water as coolant and moderator are put in perspective. The physicochemical and metallurgical properties of the materials impose thermal limits that determine the performance and the maximum power a reactor can deliver. (author)

Occupational radiation exposure at commercial nuclear power reactors, 1978

International Nuclear Information System (INIS)

Brooks, B.G.

1979-11-01

An updated compilation is presented of occupational radiation exposures at commercial nuclear power reactors for the years 1969 through 1978. Data received from the 64 light water cooled reactors (LWRs) that had completed at least one year of commercial operation as of December 31, 1978 are included. This represents an increase of seven reactors over the number contained in last year's report. The total number of personnel monitored at LWRs during 1978 increased by approximately 12% to 76,121. The number of workers that received measurable doses, however, increased only 8% to 45,978. The total collective dose for 1978 is estimated to be 31,806 man-rems, a small decrease from last year's value of 32,511, which results in the average dose per worker decreasing slightly to 0.69 rems. The average collective dose per reactor also decreased, by approximately 15%, to a value of 497 man-rems

Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

International Nuclear Information System (INIS)

Massimino, R.J.; Williams, D.A.

1983-05-01

This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core

Installation of the Light-Water Breeder Reactor at the Shippingport Atomic Power Station (LWBR Development Program)

Energy Technology Data Exchange (ETDEWEB)

Massimino, R.J.; Williams, D.A.

1983-05-01

This report summarizes the refueling operations performed to install a Light Water Breeder Reactor (LWBR) core into the existing pressurized water reactor vessel at the Shippingport Atomic Power Station. Detailed descriptions of the major installation operations (e.g., primary system preconditioning, fuel installation, pressure boundary seal welding) are included as appendices to this report; these operations are of technical interest to any reactor servicing operation, whether the reactor is a breeder or a conventional light water non-breeder core.

Trends in light water reactor dosimetry programs

International Nuclear Information System (INIS)

Rahn, F.J.; Serpan, C.Z.; Fabry, A.; McElroy, W.N.; Grundl, J.A.; Debrue, J.

1977-01-01

Dosimetry programs and techniques play an essential role in the continued assurance of the safety and reliability of components of light water reactors. Primary concern focuses on the neutron irradiation embrittlement of reactor pressure vessels and methods by which the integrity of a pressure vessel can be predicted and monitored throughout its service life. Research in these areas requires a closely coordinated program which integrates the elements of the calculational and material sciences, the development of advanced dosimetric techniques and the use of benchmarks and validation of these methods. The paper reviews the status of the various international efforts in the dosimetry area

Competition for light and water in a coupled soil-plant system

Science.gov (United States)

Manoli, Gabriele; Huang, Cheng-Wei; Bonetti, Sara; Domec, Jean-Christophe; Marani, Marco; Katul, Gabriel

2017-10-01

It is generally accepted that resource availability shapes the structure and function of many ecosystems. Within the soil-plant-atmosphere (SPA) system, resource availability fluctuates in space and time whereas access to resources by individuals is further impacted by plant-to-plant competition. Likewise, transport and transformation of resources within an individual plant is governed by numerous interacting biotic and abiotic processes. The work here explores the co-limitations on water losses and carbon uptake within the SPA arising from fluctuating resource availability and competition. In particular, the goal is to unfold the interplay between plant access and competition for water and light, as well as the impact of transport/redistribution processes on leaf-level carbon assimilation and water fluxes within forest stands. A framework is proposed that couples a three-dimensional representation of soil-root exchanges with a one-dimensional description of stem water flow and storage, canopy photosynthesis, and transpiration. The model links soil moisture redistribution, root water uptake, xylem water flow and storage, leaf potential and stomatal conductance as driven by supply and demand for water and carbon. The model is then used to investigate plant drought resilience of overstory-understory trees simultaneously competing for water and light. Simulation results reveal that understory-overstory interactions increase ecosystem resilience to drought (i.e. stand-level carbon assimilation rates and water fluxes can be sustained at lower root-zone soil water potentials). This resilience enhancement originates from reduced transpiration (due to shading) and hydraulic redistribution in soil supporting photosynthesis over prolonged periods of drought. In particular, the presence of different rooting systems generates localized hydraulic redistribution fluxes that sustain understory transpiration through overstory-understory interactions. Such complex SPA dynamics

Recent digital control and protection retrofits in power plants

International Nuclear Information System (INIS)

Fournier, R.D.; Hammer, M.; Smith, J.E.

1987-01-01

Digital computers are now being retrofitted to all types of power plants, replacing analog equipment and solving problems such as equipment obsolescence and low reliability. Three diverse examples of retrofits are presented in this paper, representing trends in man/machine interface design at an oil-fired plant, protection system in pressurized heavy-water reactors, and control systems in light water reactors (LWRs). The examples have been chosen to illustrate diverse reasons for the retrofits and the benefits derived. The cases presented report retrofits at Northern States Power's Monticello boiling water reactor, New Brunswick Electric Power Commission's (NBEPC's) Point Lepreau Nuclear Generating Station, and finally NBEPC's oil-fired plant at Courtney Bay

Future trends of light-water reactor fuel-cycle costs

International Nuclear Information System (INIS)

Tamiya, S.; Otomo, T.; Meguro, T.

1977-01-01

In past cost estimates, the main fuel-cycle components were mining and milling, uranium enrichment and fuel fabrication, and the reprocessing charge deemed to be recovered by plutonium credit. Since the oil crisis, all costs of fuel-cycle components have increased as well as construction costs of power stations. Recent analysis shows that costs in the back-end of the fuel cycle are much higher than those anticipated several years ago, although their contribution to the nuclear electricity generating cost would be small. This situation has now changed but there are still many uncertainties, i.e. regulatory requirements for reprocessing plants concerning safety, safeguards, environmental protection and high-level waste management. Thus it is more difficult to estimate overall investments and, therefore, the costs of the back-end of the fuel-cycle sector. This paper reviews the future trend of nuclear fuel-cycle costs of LWRs, based on recent analysis; and those factors which affect fuel-cycle costs are discussed. To reduce the uncertainties of the cost estimates as soon as possible, international discussion is necessary on items such as the treatment and disposal of high-level radioactive wastes, siting reprocessing plants, physical protection of plutonium, and the effects of plutonium on the environment. (author)

Development of next-generation light water reactor

International Nuclear Information System (INIS)

Ishibashi, Fumihiko; Yasuoka, Makoto

2010-01-01

The Next-Generation Light Water Reactor Development Program, a national project in Japan, was inaugurated in April 2008. The primary objective of this program is to meet the need for the replacement of existing nuclear power plants in Japan after 2030. With the aim of setting a global standard design, the reactor to be developed offers greatly improved safety, reliability, and economic efficiency through several innovative technologies, including a reactor core system with uranium enrichment of 5 to 10%, a seismic isolation system, long-life materials, advanced water chemistry, innovative construction techniques, optimized passive and active safety systems, innovative digital technologies, and so on. In the first three years, a plant design concept with these innovative features is to be established and the effectiveness of the program will be reevaluated. The major part of the program will be completed in 2015. Toshiba is actively engaged in both design studies and technology development as a founding member of this program. (author)

An investigation of differences between measured and calculated bucklings of a series of light water and heavy water moderated experimental cores

International Nuclear Information System (INIS)

Figgins, A.J.G.

1966-02-01

A series of light water and light and heavy water moderated exponential and critical experiments performed by the Babcock and Wilcox Company were analysed using the METHUSELAH programme and it was found that the calculated and measured critical bucklings differed significantly. The effect was most marked as the temperature of the moderator was raised in the light water cores where it amounted to 10 m -2 for a 200 deg. C rise above room temperature. Of this discrepancy 3 m -2 , at the most, could be explained as being caused by the experimental cores not being large enough to have a central asymptotic region, leaving an unexplained difference of 7 m -2 . It is suggested that the only region in which METHUSELAH could be usefully modified to improve this agreement is in the calculation of the resonance escape probability. The last section of the report compares the calculated and measured results obtained at room temperatures. (author)

Corrosion and stress corrosion cracking in supercritical water

Science.gov (United States)

Was, G. S.; Ampornrat, P.; Gupta, G.; Teysseyre, S.; West, E. A.; Allen, T. R.; Sridharan, K.; Tan, L.; Chen, Y.; Ren, X.; Pister, C.

2007-09-01

Supercritical water (SCW) has attracted increasing attention since SCW boiler power plants were implemented to increase the efficiency of fossil-based power plants. The SCW reactor (SCWR) design has been selected as one of the Generation IV reactor concepts because of its higher thermal efficiency and plant simplification as compared to current light water reactors (LWRs). Reactor operating conditions call for a core coolant temperature between 280 °C and 620 °C at a pressure of 25 MPa and maximum expected neutron damage levels to any replaceable or permanent core component of 15 dpa (thermal reactor design) and 100 dpa (fast reactor design). Irradiation-induced changes in microstructure (swelling, radiation-induced segregation (RIS), hardening, phase stability) and mechanical properties (strength, thermal and irradiation-induced creep, fatigue) are also major concerns. Throughout the core, corrosion, stress corrosion cracking, and the effect of irradiation on these degradation modes are critical issues. This paper reviews the current understanding of the response of candidate materials for SCWR systems, focusing on the corrosion and stress corrosion cracking response, and highlights the design trade-offs associated with certain alloy systems. Ferritic-martensitic steels generally have the best resistance to stress corrosion cracking, but suffer from the worst oxidation. Austenitic stainless steels and Ni-base alloys have better oxidation resistance but are more susceptible to stress corrosion cracking. The promise of grain boundary engineering and surface modification in addressing corrosion and stress corrosion cracking performance is discussed.

Effects of temperature and salinity on light scattering by water

Science.gov (United States)

Zhang, Xiaodong; Hu, Lianbo

2010-04-01

A theoretical model on light scattering by water was developed from the thermodynamic principles and was used to evaluate the effects of temperature and salinity. The results agreed with the measurements by Morel within 1%. The scattering increases with salinity in a non-linear manner and the empirical linear model underestimate the scattering by seawater for S < 40 psu. Seawater also exhibits an 'anomalous' scattering behavior with a minimum occurring at 24.64 °C for pure water and this minimum increases with the salinity, reaching 27.49 °C at 40 psu.

Reactor safety-a New Approach

International Nuclear Information System (INIS)

Machiels, A.J.; Marston, T.U.; Taylor, J.J.

1993-01-01

Since 1982, the U.S. utilities have been leading to an industry-wide effort to establish a technical foundation for the design of the next generation of light water reactors (LWRs) in the United States. Since 1985, the utility initiative has been effected through a major technical program managed by the Electric Power Research Institute (EPRI): the Advanced Light Water Reactor (ALWR) Program. In addition to the U.S. utility leadership and sponsorship, the ALWR Program has also greatly benefitted from the participation and sponsorship of numerous international utility companies and from the close cooperation with the U.S. Department of Energy (DOE). One of the main goals of the ALWR Program has been to develop a comprehensive set of design requirements for the advanced LWRs. The Utility Requirement Document (URD) defines the technical basis for improved and standardized future LWR designs. The URD covers the entire plant up to the grid interface. Therefore, it is the basis for an integrated plant design, i.e., nuclear steam supply system and balance of plant. It emphasizes those areas which are most important to the objective of achieving an advanced LWR that is excellent with respect to safety, performance, constructibility, and economics. There are numerous basic design policies underlying the ALWR URD. Of particular interest is the treatment of reactor safety

Comparison of two thorium fuel cycles for use in light water prebreeder/breeder reactor systems (AWBA Development Program)

International Nuclear Information System (INIS)

Merriman, F.C.; McCoy, D.F.; Boyd, W.A.; Dwyer, J.R.

1983-05-01

Light water prebreeder/breeder conceptual reactor systems have been developed which have the potential to significantly improve the fuel utilization of present generation light water reactors. The purpose of this study is to describe and compare two possible types of thorium fuel cycles for use in these light water prebreeder and breeder concepts. The two types of thorium fuel cycles basically differ in the fuel rod design used in the prebreeder cores and the uranium isotopic concentration of fuel supplied to the breeder cores

Design data and safety features of commerical nuclear power plant

International Nuclear Information System (INIS)

Heddleson, F.A.

1976-06-01

Design data, safety features, and site characteristics are summarized for 34 nuclear power units in 17 power stations in the United States. Six pages of data are presented for each plant, consisting of thermal-hydraulic and nuclear factors, containment features, emergency-core-cooling systems, site features, circulating water system data, and miscellaneous factors. An aerial perspective is also presented for each plant. This volume covers Light Water Reactors (LWRs) with dockets 50-508 through 50-549, four HTGRs--50-171, 50-267, 50-450/451, 50-463/464, the Atlantic Floating Station 50-477/478, and the Clinch River Breeder 50-537

Studies of severe accidents in light water reactors. Containment performance

International Nuclear Information System (INIS)

Hayns, M.R.; Phillips, D.W.; Young, R.L.D.

1987-01-01

The containment system of a LWR is an obvious component of the plant which performs an important safety function in preventing the release of fission products to the environment in the event of design basis accidents. With over 260 LWRs in service worldwide, and others still under construction, there is a considerable diversity of containment types and combinations of containment safeguards systems. All of these satisfy local regulatory requirements which are principally aimed at the design basis accidents, and these requirements naturally have a considerable uniformity. However, their design diversity becomes more relevant to the performance of the containment in severe accident conditions, and this aspect of containment performance is reviewed in this paper. The ability of the containment to mitigate severe accident consequences introduces the potential for accident management and recovery and this in turn points towards a range of new containment systems and concepts. PSA helps in judging these possibilities and in forming policies and procedures for accident management. It is perhaps in accident management that severe accident containment performance will be most beneficial in the future, and where additional effort in containment analysis will be focused

European community light water reactor safety research projects. Experimental issue

International Nuclear Information System (INIS)

1975-01-01

Research programs on light water reactor safety currently carried out in the European Community are presented. They cover: accident conditions (LOCA, ECCS, core meltdown, external influences, etc...), fault and accident prevention and means of mitigation, normal operation conditions, on and off site implications and equipment under severe accident conditions, and miscellaneous subjects

Light Water Reactor Sustainability Program Reactor Safety Technologies Pathway Technical Program Plan

Energy Technology Data Exchange (ETDEWEB)

Corradini, M. L. [Univ. of Wisconsin, Madison, WI (United States); Peko, D. [US Dept. of Energy, Washington, DC (United States); Farmer, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Rempe, J. [Rempe and Associates LLC, Idaho Falls, ID (United States); Humrickhouse, P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); O' Brien, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Robb, K. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gauntt, R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Osborn, D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2016-06-01

“end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.

Light Water Reactor Sustainability Program: Reactor Safety Technologies Pathway Technical Program Plan

Energy Technology Data Exchange (ETDEWEB)

Corradini, M. L. [Univ. of Wisconsin, Madison, WI (United States)

2015-06-01

“end user” of the results from this DOE-sponsored work. The response to the Fukushima accident has been global, and there is a continuing multinational interest in collaborations to better quantify accident consequences and to incorporate lessons learned from the accident. DOE will continue to seek opportunities to facilitate collaborations that are of value to the U.S. industry, particularly where the collaboration provides access to vital data from the accident or otherwise supports or leverages other important R&D work. The purpose of the Reactor Safety Technology R&D is to improve understanding of beyond design basis events and reduce uncertainty in severe accident progression, phenomenology, and outcomes using existing analytical codes and information gleaned from severe accidents, in particular the Fukushima Daiichi events. This information will be used to aid in developing mitigating strategies and improving severe accident management guidelines for the current light water reactor fleet.

Development status and application prospect of supercritical-pressure light water cooled reactor

International Nuclear Information System (INIS)

Li Manchang; Wang Mingli

2006-01-01

The Supercritical-pressure Light Water Cooled Reactor (SCWR) is selected by the Generation IV International Forum (GIF) as one of the six Generation IV nuclear systems that will be developed in the future, and it is an innovative design based on the existing technologies used in LWR and supercritical coal-fired plants. Technically, SCWR may be based on the design, construction and operation experiences in existing PWR and supercritical coal-fired plants, which means that there is no insolvable technology difficulties. Since PWR technology will be adopted in the near term and medium term projects in China, and considering the sustainable development of the technology, it is an inevitable choice to research and develop the nuclear system of supercritical light water cooled reactor. (authors)

10 CFR 50.46 - Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors.

Science.gov (United States)

2010-01-01

... light-water nuclear power reactors. 50.46 Section 50.46 Energy NUCLEAR REGULATORY COMMISSION DOMESTIC... reactors. (a)(1)(i) Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide... evaluation model. This section does not apply to a nuclear power reactor facility for which the...

Future of nuclear power in Japan - Development of next Generation LWRs

Energy Technology Data Exchange (ETDEWEB)

Yamada, Eiji; Yamamoto, T.; Kurosaki, K.; Ohga, Y.; Tsuzuki, K.; Kasai, S.; Tanaka, T.

2010-09-15

Japan's energy policies have been to decrease the oil portion and dependence on the Middle East for energy security, as well as satisfy environmental requirement. The report of 2008 targeted reducing GHG emission by 60-80% before 2050, and highlighted ''Cool Earth-Innovative Energy Technology Program'' featuring 21 innovative technologies. In this context nuclear power is expected as a core power source. In April 2008, ''Next Generation Light Water Power Reactor Development Program'' was launched with the IAE as the core organization in alliance with Japan's major vendors and in collaboration with METI and power utilities for the future of nuclear power.

Research and development of super light water reactors and super fast reactors in Japan

International Nuclear Information System (INIS)

Oka, Y.; Morooka, S.; Yamakawa, M.; Ishiwatari, Y.; Ikejiri, S.; Katsumura, Y.; Muroya, Y.; Terai, T.; Sasaki, K.; Mori, H.; Hamamoto, Y.; Okumura, K.; Kugo, T.; Nakatsuka, T.; Ezato, K.; Akasaka, N.; Hotta, A.

2011-01-01

Super Light Water Reactors (Super LWR) and Super Fast Reactors (Super FR) are the supercritical- pressure light water cooled reactors (SCWR) that are developed by the research group of University of Tokyo since 1989 and now jointly under development with the researchers of Waseda University, University of Tokyo and other organizations in Japan. The principle of the reactor concept development, the results of the past Super LWR and Super FR R&D as well as the R&D program of the Super FR second phase project are described. (author)

Fuel Summary Report: Shippingport Light Water Breeder Reactor

International Nuclear Information System (INIS)

Illum, D.B.; Olson, G.L.; McCardell, R.K.

1999-01-01

The Shippingport Light Water Breeder Reactor (LWBR) was a small water cooled, U-233/Th-232 cycle breeder reactor developed by the Pittsburgh Naval Reactors to improve utilization of the nation's nuclear fuel resources in light water reactors. The LWBR was operated at Shippingport Atomic Power Station (APS), which was a Department of Energy (DOE) (formerly Atomic Energy Commission)-owned reactor plant. Shippingport APS was the first large-scale, central-station nuclear power plant in the United States and the first plant of such size in the world operated solely to produce electric power. The Shippingport LWBR was operated successfully from 1977 to 1982 at the APS. During the five years of operation, the LWBR generated more than 29,000 effective full power hours (EFPH) of energy. After final shutdown, the 39 core modules of the LWBR were shipped to the Expended Core Facility (ECF) at Naval Reactors Facility at the Idaho National Engineering and Environmental Laboratory (INEEL). At ECF, 12 of the 39 modules were dismantled and about 1000 of more than 17,000 rods were removed from the modules of proof-of-breeding and fuel performance testing. Some of the removed rods were kept at ECF, some were sent to Argonne National Laboratory-West (ANL-W) in Idaho and some to ANL-East in Chicago for a variety of physical, chemical and radiological examinations. All rods and rod sections remaining after the experiments were shipped back to ECF, where modules and loose rods were repackaged in liners for dry storage. In a series of shipments, the liners were transported from ECF to Idaho Nuclear Technology Engineering Center (INTEC), formerly the Idaho Chemical Processing Plant (ICPP). The 47 liners containing the fully-rodded and partially-derodded core modules, the loose rods, and the rod scraps, are now stored in underground dry wells at CPP-749

Occupational radiation exposure at commercial nuclear power reactors, 1981. Annual report

International Nuclear Information System (INIS)

Brooks, B.G.

1982-11-01

This report presents an updated compilation of occupational radiation exposures at commercial nuclear power reactors for the years 1969 through 1981. This year's report contains data received from the 70 light water cooled reactors (LWRs) and one high temperature gas cooled reactor that had been declared to be in commercial operation for at least one full year as of December 31, 1981. This represents an increase of two reactors over the number contained in last year's report. The total number of personnel monitored at LWRs in 1981 was 124,504, a slight decrease from that found in 1980. The number of workers that received measurable doses during 1981 was 82,183 which is about 2000 more than that found in 1980. The total collective dose at LWRs for 1981 is estimated to be 54,142 man-rems, which is only about 350 man-rems more than that reported in 1980. The report also presents a summary and some analyses of the exposure data contained in the termination reports that have been submitted by nuclear power licensees to the Commission pursuant to 10 CFR Section 20.408. As of December 31, 1981, personal identification and exposure information had been collected and computerized for some 210,000 of these terminating reactor personnel

Status and topics of thermal-hydraulic analysis for next-generation LWRs with passive safety systems

International Nuclear Information System (INIS)

Aritomi, Masanori; Ohnuki, Akira; Arai, Kenji; Kikuta, Michitaka; Yonomoto, Taisuke; Araya, Fumimasa; Akimoto, Hajime

1999-01-01

For increasing of electric power demand and reducing of carbon dioxide exhaust in the 21st century, studies of the next-generation light water reactor (LWR) with passive safety systems are developing in the world: AP-600 (by Westing House Co.); SBWR (by General Electric Co.); SWR1000 (by Siemens Co.); NP21 (by Mitsubishi Heavy Industry Co., et al.); JPSR (by JAERI). The passive equipment using natural circulation and natural convection are installed in the passive safety system, instead of active safety equipment, such as pumps, etc. It remains still as a important issue, however, to verify the reliability on the functions of the passive equipment, since that the driving forces of the passive equipment are small at comparison with the active safety equipment. The various subjects of thermal-hydraulic analysis for the next-generation light water reactors, such as temperature stratification in the passive safety systems, vapor condensation in the mixture of non-condensable gases and the interactions of the passive safety system with the primary cooling system, are illustrated and discussed in the paper. (M. Suetake)

Tritium formation and elimination in light-water electronuclear plants

International Nuclear Information System (INIS)

Dolle, L.; Bazin, J.

1977-01-01

In light-water reactors, the tritium balance should be considered from both the working constraint and environmental pollution aspects. In light-water electronuclear stations with pressurized reactors using boric acid in solution for reactivity control, the amounts of tritium formed in the primary circuit are worthy of note. The estimations concerning the tritium production in a hypothetical 1000 MWe reactor are discussed. In the tritium build-up, the part which takes the tritium formed by fission in the fuel, owing to diffusion through cladding, is still difficult to estimate. The tritium balance in different working nuclear power stations are consequently of interest. But the tritium produced by ternary fission in the fuel is always much more abundant, and remains almost entirely confined in the uranium oxide if the fuel is clad with zircaloy. The annual quantity stored in the fuel elements is more than 20 times larger than that of the built up free tritium in the primary circuit water of a reactor. It reaches about 12,400 Ci in the hypothetical reactor. In the presently operated reprocessing plants, tritium is all going over in the effluents, and is almost entirely released in the environment. Taking into account the increasing quantities of high irradiated fuel to be reprocessed, it seems necessary to develop separation processes. Development work and tests have been achieved jointly by CEA and SAINT-GOBAIN TECHNIQUES NOUVELLES in order to: contain the tritium in the high activity part of the plant; and keep small the tritiated effluent volume, about 300 liters per ton of reprocessed uranium. It is then possible to envisage a storage for decay of isotopic separation processes. Such separation processes have been estimated by CEA assuming a daily output of 1500 liters of water containing 2,3 Ci.1 -1 of tritium, the desired decontamination factor being 100 [fr

Dynamic Response of Plant Chlorophyll Fluorescence to Light, Water and Nutrient Availability

Science.gov (United States)

Cendrero Mateo, M. D. P.; Moran, S. M.; Porcar-Castell, A.; Carmo-Silva, A. E.; Papuga, S. A.; Matveeva, M.; Wieneke, S.; Rascher, U.

2014-12-01

Photosynthesis is the most important exchange process of CO2 between the atmosphere and the land-surface. Spatial and temporal patterns of photosynthesis depend on dynamic plant-specific adaptation strategies to highly variable environmental conditions e.g. light, water, and nutrient availability. Chlorophyll fluorescence (ChF) has been proposed as a direct indicator of photosynthesis, and several studies have demonstrated its relationship with vegetation functioning at leaf and canopy level. In this study, two overarching questions about ChF were addressed: Q1) How water, nutrient and ambient light conditions determine the relationships between photosynthesis and ChF? Which is the optimum irradiance level for detecting water and nutrient deficit conditions with ChF?; Q2) What is the seasonal relationship between photosynthesis and ChF when nitrogen is the limiting factor? The results of this study indicated that when the differences between treatments (water or nitrogen) drive the relationship between photosynthesis and ChF, ChF has a direct relationship with photosynthesis. This study demonstrates that the light level at which plants were grown was optimum for detecting water and nutrient deficit with ChF. Further, the seasonal relation between photosynthesis and ChF with nitrogen stress was not a simple linear function due to the complex physiological relation between photosynthesis and ChF. Our study showed that at times in the season when nitrogen was sufficient and photosynthesis was highest, ChF decreased because these two processes compete for available energy. The results from this study demonstrated that ChF is a reliable indicator of plant stress and has great potential as a tool for better understand where, when, and how CO2 is exchanged between the land and atmosphere.

Current and anticipated uses of thermal hydraulic codes at the Japan Atomic Energy Research Institute

Energy Technology Data Exchange (ETDEWEB)

Akimoto, Hajime; Kukita; Ohnuki, Akira [Japan Atomic Energy Research Institute, Ibaraki (Japan)

1997-07-01

The Japan Atomic Energy Research Institute (JAERI) is conducting several research programs related to thermal-hydraulic and neutronic behavior of light water reactors (LWRs). These include LWR safety research projects, which are conducted in accordance with the Nuclear Safety Commission`s research plan, and reactor engineering projects for the development of innovative reactor designs or core/fuel designs. Thermal-hydraulic and neutronic codes are used for various purposes including experimental analysis, nuclear power plant (NPP) safety analysis, and design assessment.

Investigation of valve failure problems in LWR power plants

Energy Technology Data Exchange (ETDEWEB)

None

1980-04-01

An analysis of component failures from information in the computerized Nuclear Safety Information Center (NSIC) data bank shows that for both PWR and BWR plants the component category most responsible for approximately 19.3% of light water reactor (LWR) power plant shutdowns. This investigation by Burns and Roe, Inc. shows that the greatest cause of shutdowns in LWRs due to valve failures is leakage from valve stem packing. Both BWR plants and PWR plants have stem leakage problems (BWRs, 21% and PWRs, 34%).

Advanced fuel cycles options for LWRs and IMF benchmark definition

International Nuclear Information System (INIS)

Breza, J.; Darilek, P.; Necas, V.

2008-01-01

In the paper, different advanced nuclear fuel cycles including thorium-based fuel and inert-matrix fuel are examined under light water reactor conditions, especially VVER-440, and compared. Two investigated thorium based fuels include one solely plutonium-thorium based fuel and the second one plutonium-thorium based fuel with initial uranium content. Both of them are used to carry and burn or transmute plutonium created in the classical UOX cycle. The inert-matrix fuel consist of plutonium and minor actinides separated from spent UOX fuel fixed in Yttria-stabilised zirconia matrix. The article shows analysed fuel cycles and their short description. The conclusion is concentrated on the rate of Pu transmutation and Pu with minor actinides cumulating in the spent advanced thorium fuel and its comparison to UOX open fuel cycle. Definition of IMF benchmark based on presented scenario is given. (authors)

Light Refraction by Water as a Rationale for the Poggendorff Illusion.

Science.gov (United States)

Bozhevolnyi, Sergey I

2016-08-24

The Poggendorff illusion in its classical form of parallel lines interrupting a transversal is viewed from the perspective of being related to the everyday experience of observing the light refraction in water. It is argued that if one considers a transversal to be a light ray in air and the parallel lines to form an occluding strip of a medium with the refractive index being between that of air and water, then one should be able to account, both qualitatively and quantitatively, for most of the features associated with the Poggendorff illusion. Statistical treatment of the visual experiments conducted with seven participants, each analyzing 50 configurations having different intercepting angles and strip widths, resulted in the effective refractive index of the occluding strip N = 1.13 ± 0.15, which is sufficiently close to the average (between that of water and air) refractive index of ∼1.17. It is further argued that the same mechanism can also be employed to account for many variants of the Poggendorff illusion, including the corner-Poggendorff pattern, as well as for the Hering illusion. © The Author(s) 2016.

Water chemistry and materials degradation in LWR'S

International Nuclear Information System (INIS)

Haenninen, H.; Toerroenen, K.; Aaltonen, P.

1994-01-01

Water chemistry plays a major role in corrosion, in erosion corrosion and in activity transport in NPPs; it impacts upon the operational safety of LWRs in two main ways: integrity of pressure boundary materials and activity transport and out-of-core radiation fields. A good control of water chemistry can significantly reduce these problems and improve plant safety, but economic pressures are leading to more rigorous operating conditions: fuel burnups are to be increased, higher efficiencies are to be achieved by running at higher temperatures and plant lifetimes are to be extended. Typical water chemistry specifications used in PWR and BWR plants are presented and the chemistry optimization is discussed. The complex interplay of metallurgical, mechanical and environmental factors in environmental sensitive cracking is shown, with details on studies for carbon steels, stainless steels and nickel base alloys. 20 refs., 8 figs., 4 tabs

Light-water reactor pressure vessel surveillance standards

International Nuclear Information System (INIS)

Anon.

1981-01-01

The master matrix standard describes a series of standard practices, guides, and methods for the prediction of neutron-induced changes in light-water reactor (LWR) pressure vessel steels throughout a pressure vessel's service life. Some of these are existing American Society for Testing and Materials (ASTM) standards, some are ASTM standards that have been modified, and some are newly proposed ASTM standards. The current (1) scope, (2) areas of application, (3) interrelationships, and (4) status and time table of development, improvement, validation, and calibration for a series of 16 ASTM standards are defined. The standard also includes a discussion of LWR pressure vessel surveillance - justification, requirements, and status of work

Fuel utilization potential in light water reactors with once-through fuel irradiation (AWBA Development Program)

International Nuclear Information System (INIS)

Rampolla, D.S.; Conley, G.H.; Candelore, N.R.; Cowell, G.K.; Estes, G.P.; Flanery, B.K.; Duncombe, E.; Dunyak, J.; Satterwhite, D.G.

1979-07-01

Current commercial light water reactor cores operate without recylce of fuel, on a once-through fuel cycle. To help conserve the limited nuclear fuel resources, there is interest in increasing the energy yield and, hence, fuel utilization from once-through fuel irradiation. This report evaluates the potential increase in fuel utilization of light water reactor cores operating on a once-through cycle assuming 0.2% enrichment plant tails assay. This evaluation is based on a large number of survey calculations using techniques which were verified by more detailed calculations of several core concepts. It is concluded that the maximum fuel utilization which could be achieved by practical once-through pressurized light water reactor cores with either uranium or thorium is about 17 MWYth/ST U 3 O 8 (Megawatt Years Thermal per Short Ton of U 3 O 8 ). This is about 50% higher than that of current commercial light water reactor cores. Achievement of this increased fuel utilization would require average fuel burnup beyond 50,000 MWD/MT and incorporation of the following design features to reduce parasitic losses of neutrons: reflector blankets to utilize neutrons that would otherwise leak out of the core; fuel management practices in which a smaller fraction of the core is replaced at each refueling; and neutron economic reactivity control, such as movable fuel control rather than soluble boron control. For a hypothetical situation in which all neutron leakage and parasitic losses are eliminated and fuel depletion is not limited by design considerations, a maximum fuel utilization of about 20 MWYth/ST U 3 O 8 is calculated for either uranium or thorium. It is concluded that fuel utilization for comparable reactor designs is better with uranium fuel than with thorium fuel for average fuel depletions of 30,000 to 35,000 MWD/MT which are characteristic of present light water reactor cores

Leaf physiology and biomass allocation of backcross hybrid American chestnut (Castanea dentata) seedlings in response to light and water availability.

Science.gov (United States)

Brown, Caleb E; Mickelbart, Michael V; Jacobs, Douglass F

2014-12-01

Partial canopy cover promotes regeneration of many temperate forest trees, but the consequences of shading on seedling drought resistance are unclear. Reintroduction of blight-resistant American chestnut (Castanea dentata (Marsh.) Borkh.) into eastern North American forests will often occur on water-limited sites and under partial canopy cover. We measured leaf pre-dawn water potential (Ψpd), leaf gas exchange, and growth and biomass allocation of backcross hybrid American chestnut seedlings from three orchard sources grown under different light intensities (76, 26 and 8% full photosynthetically active radiation (PAR)) and subjected to well-watered or mid-season water-stressed conditions. Seedlings in the water-stress treatment were returned to well-watered conditions after wilting to examine recovery. Seedlings growing under medium- and high-light conditions wilted at lower leaf Ψpd than low-light seedlings. Recovery of net photosynthesis (Anet) and stomatal conductance (gs) was greater in low and medium light than in high light. Seed source did not affect the response to water stress or light level in most cases. Between 26 and 8% full PAR, light became limiting to the extent that the effects of water stress had no impact on some growth and morphological traits. We conclude that positive and negative aspects of shading on seedling drought tolerance and recovery are not mutually exclusive. Partial shade may help American chestnut tolerate drought during early establishment through effects on physiological conditioning. © The Author 2014. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

Technology programs in support of advanced light water reactor plants: Construction

International Nuclear Information System (INIS)

Eichen, E.P.

1989-10-01

Stone ampersand Webster Engineering Corporation (SWEC) is conducting several independent, yet interrelated, studies of light water reactor plants to improve constructibility and quality, to reduce costs and schedule duration, and to simplify design. This document discusses construction approaches. 77 refs., 5 figs., 6 tabs

Development of Radiation-Resistant In-Water Wireless Transmission System Using Light Emitting Diodes and Photo Diodes

Science.gov (United States)

Takeuchi, T.; Shibata, H.; Otsuka, N.; Uehara, T.; Tsuchiya, K.; Shibagaki, T.; Komanome, H.

2016-10-01

Several kinds of commercially available light emitting diodes (LED) and photo diodes (PD) were irradiated with 60Co gamma ray up to 1 MGy for development of a radiation-resistant in-water wireless transmission system using visible light. The lens parts of the LEDs turned brown by the irradiation and their colors became dark with the absorbed dose. The total luminous fluxes decreased with the absorbed dose and the LED with shorter emission wavelength had the higher decrease rate. Meanwhile, the current-voltage characteristics hardly changed. These results indicate that the decreases of the total luminous flux of the LEDs were mainly caused not by the degradation of the semiconductor parts but by the coloring of the lens parts by the irradiation. On the other hand, the light sensitivities of the PDs decreased with the absorbed dose. The PDs with the window part which turned a darker color had the higher decrease rate. These results indicate that the decreases of light sensitivities of the PDs were also mainly caused by the coloring of the resin parts by the irradiation. If the wireless transmission is performed using the candidate LED and PD between 5 meters in water, using a few LEDs and PDs, the PD's output current generated by the emission light of the LED is estimated to be detectable even considering the effects of the absorption of the light in water and the increased dark current by the irradiation. Therefore, a radiation resistant in-water transmission system can be constructed using commercially available LEDs and PDs in principle.

A light water excess heat reaction suggests that cold fusion may be alkali-hydrogen fusion

International Nuclear Information System (INIS)

Bush, R.T.

1992-01-01

This paper reports that Mills and Kneizys presented data in support of a light water excess heat reaction obtained with an electrolytic cell highly reminiscent of the Fleischmann-Pons cold fusion cell. The claim of Mills and Kneizys that their excess heat reaction can be explained on the basis of a novel chemistry, which supposedly also explains cold fusion, is rejected in favor of their reaction being, instead, a light water cold fusion reaction. It is the first known light water cold fusion reaction to exhibit excess heat, it may serve as a prototype to expand our understanding of cold fusion. From this new reactions are deduced, including those common to past cold fusion studies. This broader pattern of nuclear reactions is typically seen to involve a fusion of the nuclides of the alkali atoms with the simplest of the alkali-type nuclides, namely, protons, deuterons, and tritons. Thus, the term alkali-hydrogen fusion seems appropriate for this new type of reaction with three subclasses: alkali-hydrogen fusion, alkali-deuterium fusion, and alkali-tritium fusion. A new three-dimensional transmission resonance model (TRM) is sketched. Finally, preliminary experimental evidence in support of the hypothesis of a light water nuclear reaction and alkali-hydrogen fusion is reported. Evidence is presented that appears to strongly implicate the transmission resonance phenomenon of the new TRM

A new device for acquiring ground truth on the absorption of light by turbid waters

Science.gov (United States)

Klemas, V. (Principal Investigator); Srna, R.; Treasure, W.

1974-01-01

The author has identified the following significant results. A new device, called a Spectral Attenuation Board, has been designed and tested, which enables ERTS-1 sea truth collection teams to monitor the attenuation depths of three colors continuously, as the board is being towed behind a boat. The device consists of a 1.2 x 1.2 meter flat board held below the surface of the water at a fixed angle to the surface of the water. A camera mounted above the water takes photographs of the board. The resulting film image is analyzed by a micro-densitometer trace along the descending portion of the board. This yields information on the rate of attenuation of light penetrating the water column and the Secchi depth. Red and green stripes were painted on the white board to approximate band 4 and band 5 of the ERTS MSS so that information on the rate of light absorption by the water column of light in these regions of the visible spectrum could be concurrently measured. It was found that information from a red, green, and white stripe may serve to fingerprint the composition of the water mass. A number of these devices, when automated, could also be distributed over a large region to provide a cheap method of obtaining valuable satellite ground truth data at present time intervals.

Automated ultrasonic examination of light water reactor systems

International Nuclear Information System (INIS)

Walter, J.H.

1975-01-01

An automated ultrasonic examination system has been developed to meet the pre- and inservice inspection requirements of light water reactors. This system features remotely-controlled travelling instrument carriers, computerized collection and storage or inspection data in a manner providing real time comparison against code standards, and computer control over the positioning of the instrument carriers to provide precise location data. The system is currently being utilized in the field for a variety of reactor inspections. The principal features of the system and the recent inspection experience are discussed. (author)

The industry/EPRI advanced light water reactor program

International Nuclear Information System (INIS)

Stahlkopf, K.E.; Noble, D.M.; Sugnet, W.R.; Bilan, W.J.

1986-01-01

For the United States nuclear power industry to remain viable, it must be prepared to meet the expected need for new generating capacity in the late 1990s with an improved reactor system. The best hope of meeting this requirement is with evolutionary changes in current LWR systems through system simplification and reevaluation of safety and operational design margins. The grid characteristics and the difficulty in raising capital for large projects indicate that smaller light water reactors (400 to 600 MWe) may play an important role the next generation

Light water ultra-safe plant concept

International Nuclear Information System (INIS)

Klevans, E.

1989-01-01

Since the accident at Three Mile Island (TMI), Penn State Nuclear Engineering Department Faculty and Staff have considered various methods to improve already safe reactor designs and public perception of the safety of Nuclear Power. During 1987 and 1988, the Department of Energy provided funds to the Nuclear Engineering Department at Penn State to investigate a plant reconfiguration originated by M.A. Schultz called ''The Light Water Ultra-Safe Plant Concept''. This report presents a final summary of the project with references to several masters' theses and addendum reports for further detail. The two year research effort included design verification with detailed computer simulation of: (a) normal operation characteristics of the unique pressurizing concept, (b) severe transients without loss of coolant, (c) combined primary and secondary system modeling, and (d) small break and large break loss of coolant accidents. Other studies included safety analysis, low power density core design, and control system design to greatly simplify the control room and required operator responses to plant upset conditions. The overall conclusion is that a reconfigured pressurized water reactor can achieve real and perceived safety improvements. Additionally, control system research to produce greatly simplified control rooms and operator requirements should be continued in future projects

Technology programs in support of advanced light water reactor plants: Construction

International Nuclear Information System (INIS)

Eichen, E.P.

1987-12-01

Stone ampersand Webster Engineering Corporation (SWEC) is conducting several independent, yet interrelated, studies of light water reactor plants to improve constructibility and quality, to reduce costs and schedule durations, and to simplify design. This document discusses successes and problems in construction. 49 refs., 16 figs., 8 tabs

Technical specification: Mixed-oxide pellets for the light-water reactor irradiation demonstration test

International Nuclear Information System (INIS)

Cowell, B.S.

1997-06-01

This technical specification is a Level 2 Document as defined in the Fissile Materials Disposition Program Light-Water Reactor Mixed-oxide Fuel Irradiation Test Project Plan. It is patterned after the pellet specification that was prepared by Atomic Energy of Canada, Limited, for use by Los Alamos National Laboratory in fabrication of the test fuel for the Parallex Project, adjusted as necessary to reflect the differences between the Canadian uranium-deuterium reactor and light-water reactor fuels. This specification and the associated engineering drawing are to be utilized only for preparation of test fuel as outlined in the accompanying Request for Quotation and for additional testing as directed by Oak Ridge National Laboratory or the Department of Energy

Photocatalytic hydrogen generation from water under visible light using core/shell nano-catalysts.

Science.gov (United States)

Wang, X; Shih, K; Li, X Y

2010-01-01

A microemulsion technique was employed to synthesize nano-sized photocatalysts with a core (CdS)/shell (ZnS) structure. The primary particles of the photocatalysts were around 10 nm, and the mean size of the catalyst clusters in water was about 100 nm. The band gaps of the catalysts ranged from 2.25 to 2.46 eV. The experiments of photocatalytic H(2) generation showed that the catalysts (CdS)(x)/(ZnS)(1-x) with x ranging from 0.1 to 1 were able to produce hydrogen from water photolysis under visible light. The catalyst with x=0.9 had the highest rate of hydrogen production. The catalyst loading density also influenced the photo-hydrogen production rate, and the best catalyst concentration in water was 1 g L(-1). The stability of the nano-catalysts in terms of size, morphology and activity was satisfactory during an extended test period for a specific hydrogen production rate of 2.38 mmol g(-1) L(-1) h(-1) and a quantum yield of 16.1% under visible light (165 W Xe lamp, lambda>420 nm). The results demonstrate that the (CdS)/(ZnS) core/shell nano-particles are a novel photo-catalyst for renewable hydrogen generation from water under visible light. This is attributable to the large band-gap ZnS shell that separates the electron/hole pairs generated by the CdS core and hence reduces their recombinations.

Core Design and Deployment Strategy of Heavy Water Cooled Sustainable Thorium Reactor

Directory of Open Access Journals (Sweden)

Naoyuki Takaki

2012-08-01

Full Text Available Our previous studies on water cooled thorium breeder reactor based on matured pressurized water reactor (PWR plant technology concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array and using heavy water as coolant is appropriate for achieving better breeding performance and higher burn-up simultaneously [1–6]. One optimum core that produces 3.5 GW thermal energy using Th-²³³U oxide fuel shows a breeding ratio of 1.07 and averaged burn-up of about 80 GWd/t with long cycle length of 1300 days. The moderator to fuel volume ratio is 0.6 and required enrichment of ²³³U for the fresh fuel is about 7%. The coolant reactivity coefficient is negative during all cycles despite it being a large scale breeder reactor. In order to introduce this sustainable thorium reactor, three-step deployment scenario, with intermediate transition phase between current light water reactor (LWR phase and future sustainer phase, is proposed. Both in transition phase and sustainer phase, almost the same core design can be applicable only by changing fissile materials mixed with thorium from plutonium to ²³³U with slight modification in the fuel assembly design. Assuming total capacity of 60 GWe in current LWR phase and reprocessing capacity of 800 ton/y with further extensions to 1600 ton/y, all LWRs will be replaced by heavy water cooled thorium reactors within about one century then thorium reactors will be kept operational owing to its potential to sustain fissile fuels while reprocessing all spent fuels until exhaustion of massive thorium resource.

End-of-life destructive examination of light water breeder reactor fuel rods (LWBR Development Program)

International Nuclear Information System (INIS)

Richardson, K.D.

1987-10-01

Destructive examination of 12 representative Light Water Breeder Reactor fuel rods was performed following successful operation in the Shippingport Atomic Power Station for 29,047 effective full power hours, about five years. Light Water Breeder Reactor fuel rods were unique in that the thorium oxide and uranium-233 oxide fuel was contained within Zircaloy-4 cladding. Destructive examinations included analysis of released fission gas; chemical analysis of the fuel to determine depletion, iodine, and cesium levels; chemical analysis of the cladding to determine hydrogen, iodine, and cesium levels; metallographic examination of the cladding, fuel, and other rod components to determine microstructural features and cladding corrosion features; and tensile testing of the irradiated cladding to determine mechanical strength. The examinations confirmed that Light Water Breeder Reactor fuel rod performance was excellent. No evidence of fuel rod failure was observed, and the fuel operating temperature was low (below 2580 0 F at which an increased percentage of fission gas is released). 21 refs., 80 figs., 20 tabs

Isotope effect in heavy/light water suspensions of optically active gold nanoparticles

Science.gov (United States)

Kutsenko, V. Y.; Artykulnyi, O. P.; Petrenko, V. I.; Avdeev, M. V.; Marchenko, O. A.; Bulavin, L. A.; Snegir, S. V.

2018-04-01

Aqueous suspensions of optically active gold nanoparticles coated with trisodium citrate were synthesized in light (H2O) water and mixture of light and heavy (H2O/D2O) water using the modified Turkevich protocol. The objective of the paper was to verify sensitivity of neutron scattering methods (in particular, neutron reflectometry) to the potential isotope H/D substitution in the stabilizing organic shell around particles in colloidal solutions. First, the isotope effect was studied with respect to the changes in the structural properties of metal particles (size, shape, crystalline morphology) in solutions by electron microscopy including high-resolution transmission electron microscopy from dried systems. The structural factors determining the variation in the adsorption spectra in addition to the change in the optical properties of surrounding medium were discussed. Then, neutron reflectometry was applied to the layered nanoparticles anchored on a silicon wafer via 3-aminopropyltriethoxysilane molecules to reveal the presence of deuterated water molecules in the shell presumably formed by citrate molecules around the metallic core.

A review of Zircaloy fuel cladding behavior in a loss-of-coolant accident

International Nuclear Information System (INIS)

Erbacher, F.J.; Leistikow, S.

1985-09-01

The paper reviews the state-of-the-art experimental work performed in several countries with respect to the acceptance criteria established for emergency core cooling (ECC) in a loss-of-coolant accident (LOGA) of light water reactors (LWRs). It covers in detail oxidation, embrittlement, plastic deformation and coolability of deformed rod bundles. The main test results are discussed on the basis of research work performed at the Karlsruhe Nuclear Research Center (KfK) within the framework of the Nuclear Safety Project (PNS) and reference is made to test data obtained in other countries. The conclusion reached in the paper is that the major mechanisms and consequences of oxidation, deformation and emergency core cooling are sufficiently investigated in order to provide a reliable data base for safety assessments and licensing of LWRs. All test data prove that the ECC-criteria are conservative and that the coolability of an LWR and the public safety can be maintained in a LOCA. (orig.) [de

Disinfection of Spacecraft Potable Water Systems by Photocatalytic Oxidation Using UV-A Light Emitting Diodes

Science.gov (United States)

Birmele, Michele N.; O'Neal, Jeremy A.; Roberts, Michael S.

2011-01-01

Ultraviolet (UV) light has long been used in terrestrial water treatment systems for photodisinfection and the removal of organic compounds by several processes including photoadsorption, photolysis, and photocatalytic oxidation/reduction. Despite its effectiveness for water treatment, UV has not been explored for spacecraft applications because of concerns about the safety and reliability of mercury-containing UV lamps. However, recent advances in ultraviolet light emitting diodes (UV LEDs) have enabled the utilization of nanomaterials that possess the appropriate optical properties for the manufacture of LEDs capable of producing monochromatic light at germicidal wavelengths. This report describes the testing of a commercial-off-the-shelf, high power Nichia UV-A LED (250mW A365nnJ for the excitation of titanium dioxide as a point-of-use (POD) disinfection device in a potable water system. The combination of an immobilized, high surface area photocatalyst with a UV-A LED is promising for potable water system disinfection since toxic chemicals and resupply requirements are reduced. No additional consumables like chemical biocides, absorption columns, or filters are required to disinfect and/or remove potentially toxic disinfectants from the potable water prior to use. Experiments were conducted in a static test stand consisting of a polypropylene microtiter plate containing 3mm glass balls coated with titanium dioxide. Wells filled with water were exposed to ultraviolet light from an actively-cooled UV-A LED positioned above each well and inoculated with six individual challenge microorganisms recovered from the International Space Station (ISS): Burkholderia cepacia, Cupriavidus metallidurans, Methylobacterium fujisawaense, Pseudomonas aeruginosa, Sphingomonas paucimobilis and Wautersia basilensis. Exposure to the Nichia UV-A LED with photocatalytic oxidation resulted in a complete (>7-log) reduction of each challenge bacteria population in UV-A LEDs and semi

Nuclear Data Libraries for Hydrogen in Light Water Ice

International Nuclear Information System (INIS)

Torres, L; Gillette, V.H

2000-01-01

Nuclear data libraries were produced for hydrogen (H) in light water ice at different temperatures, 20, 30, 50, 77, 112, 180, 230 K.These libraries were produced using the NJOY nuclear data processing system.With this code we produce pointwise cross sections and related quantities, in the ENDF format, and in the ACE format for MCNP.Experimental neutron spectra at such temperatures were compared with MCNP4B simulations, based on the locally produced libraries, leading to satisfactory results

Intrinsic climate dependency of ecosystem light and water-use-efficiencies across Australian biomes

International Nuclear Information System (INIS)

Shi, Hao; Li, Longhui; Eamus, Derek; Cleverly, James; Huete, Alfredo; Yu, Qiang; Beringer, Jason; Van Gorsel, Eva; Hutley, Lindsay

2014-01-01

The sensitivity of ecosystem gross primary production (GPP) to availability of water and photosynthetically active radiation (PAR) differs among biomes. Here we investigated variations of ecosystem light-use-efficiency (eLUE: GPP/PAR) and water-use-efficiency (eWUE: GPP/evapotranspiration) among seven Australian eddy covariance sites with differing annual precipitation, species composition and temperature. Changes to both eLUE and eWUE were primarily correlated with atmospheric vapor pressure deficit (VPD) at multiple temporal scales across biomes, with minor additional correlations observed with soil moisture and temperature. The effects of leaf area index on eLUE and eWUE were also relatively weak compared to VPD, indicating an intrinsic dependency of eLUE and eWUE on climate. Additionally, eLUE and eWUE were statistically different for biomes between summer and winter, except eWUE for savannas and the grassland. These findings will improve our understanding of how light- and water-use traits in Australian ecosystems may respond to climate change. (letter)

Navigation by light polarization in clear and turbid waters

Science.gov (United States)

Lerner, Amit; Sabbah, Shai; Erlick, Carynelisa; Shashar, Nadav

2011-01-01

Certain terrestrial animals use sky polarization for navigation. Certain aquatic species have also been shown to orient according to a polarization stimulus, but the correlation between underwater polarization and Sun position and hence the ability to use underwater polarization as a compass for navigation is still under debate. To examine this issue, we use theoretical equations for per cent polarization and electric vector (e-vector) orientation that account for the position of the Sun, refraction at the air–water interface and Rayleigh single scattering. The polarization patterns predicted by these theoretical equations are compared with measurements conducted in clear and semi-turbid coastal sea waters at 2 m and 5 m depth over sea floors of 6 m and 28 m depth. We find that the per cent polarization is correlated with the Sun's elevation only in clear waters. We furthermore find that the maximum value of the e-vector orientation angle equals the angle of refraction only in clear waters, in the horizontal viewing direction, over the deeper sea floor. We conclude that navigation by use of underwater polarization is possible under restricted conditions, i.e. in clear waters, primarily near the horizontal viewing direction, and in locations where the sea floor has limited effects on the light's polarization. PMID:21282170

Fuel qualification issues and strategies for reactor-based surplus plutonium disposition

International Nuclear Information System (INIS)

Cowell, B.S.; Copeland, G.L.; Moses, D.L.

1997-08-01

The Department of Energy (DOE) has proposed irradiation of mixed-oxide (MOX) fuel in existing commercial reactors as a disposition method for surplus plutonium from the weapons program. The burning of MOX fuel in reactors is supported by an extensive technology base; however, the infrastructure required to implement reactor-based plutonium disposition does not exist domestically. This report identifies and examines the actions required to qualify and license weapons-grade (WG) plutonium-based MOX fuels for use in domestic commercial light-water reactors (LWRs)

Energy Technology Division research summary 2004

International Nuclear Information System (INIS)

Poeppel, R. B.; Shack, W. J.

2004-01-01

The Energy Technology (ET) Division provides materials and engineering technology support to a wide range of programs important to the US Department of Energy (DOE). The Division's capabilities are generally applied to technical issues associated with energy systems, biomedical engineering, transportation, and homeland security. Research related to the operational safety of commercial light water nuclear reactors (LWRs) for the US Nuclear Regulatory Commission (NRC) remains another significant area of interest for the Division. The pie chart below summarizes the ET sources of funding for FY 2004

Safety of next generation power reactors

International Nuclear Information System (INIS)

Anon.

1988-01-01

This book is organized under the following headings: Future needs of utilities regulators, government, and other energy users, PRA and reliability, LMR concepts, LWR design, Advanced reactor technology, What the industry can deliver: advanced LWRs, High temperature gas-cooled reactors, LMR whole-core experiments, Advanced LWR concepts, LWR technology, Forum: public perceptions, What the industry can deliver: LMRs and HTGRs, Criteria and licensing, LMR modeling, Light water reactor thermal-hydraulics, LMR technology, Working together to revitalize nuclear power, Appendix A, luncheon address, Appendix B, banquet address

Issues affecting advanced passive light-water reactor safety analysis

International Nuclear Information System (INIS)

Beelman, R.J.; Fletcher, C.D.; Modro, S.M.

1992-01-01

Next generation commercial reactor designs emphasize enhanced safety through improved safety system reliability and performance by means of system simplification and reliance on immutable natural forces for system operation. Simulating the performance of these safety systems will be central to analytical safety evaluation of advanced passive reactor designs. Yet the characteristically small driving forces of these safety systems pose challenging computational problems to current thermal-hydraulic systems analysis codes. Additionally, the safety systems generally interact closely with one another, requiring accurate, integrated simulation of the nuclear steam supply system, engineered safeguards and containment. Furthermore, numerical safety analysis of these advanced passive reactor designs wig necessitate simulation of long-duration, slowly-developing transients compared with current reactor designs. The composite effects of small computational inaccuracies on induced system interactions and perturbations over long periods may well lead to predicted results which are significantly different than would otherwise be expected or might actually occur. Comparisons between the engineered safety features of competing US advanced light water reactor designs and analogous present day reactor designs are examined relative to the adequacy of existing thermal-hydraulic safety codes in predicting the mechanisms of passive safety. Areas where existing codes might require modification, extension or assessment relative to passive safety designs are identified. Conclusions concerning the applicability of these codes to advanced passive light water reactor safety analysis are presented

Spent fuel data base: commercial light water reactors

International Nuclear Information System (INIS)

Hauf, M.J.; Kniazewycz, B.G.

1979-12-01

As a consequence of this country's non-proliferation policy, the reprocessing of spent nuclear fuel has been delayed indefinitely. This has resulted in spent light water reactor (LWR) fuel being considered as a potential waste form for disposal. Since the Nuclear Regulatory Commission (NRC) is currently developing methodologies for use in the regulation of the management and disposal of high-level and transuranic wastes, a comprehensive data base describing LWR fuel technology must be compiled. This document provides that technology baseline and, as such, will support the development of those evaluation standards and criteria applicable to spent nuclear fuel

Technology programs in support of advanced light water reactor plants: Construction

International Nuclear Information System (INIS)

Eichen, E.P.

1989-01-01

Under Contract No. AC03-86SF16565, Stone ampersand Webster Engineering Corporation (SWEC) is conducting several independent, yet interrelated, studies of light water reactor plants to improve constructibility and quality, to reduce costs and schedule durations, and to simplify design. This document discusses design requirements. 36 refs., 57 figs., 56 tabs

LWR-WIMS, a computer code for light water reactor lattice calculations

International Nuclear Information System (INIS)

Halsall, M.J.

1982-06-01

LMR-WIMS is a comprehensive scheme of computation for studying the reactor physics aspects and burnup behaviour of typical lattices of light water reactors. This report describes the physics methods that have been incorporated in the code, and the modifications that have been made since the code was issued in 1972. (U.K.)

A method and algorithm for correlating scattered light and suspended particles in polluted water

International Nuclear Information System (INIS)

Sami Gumaan Daraigan; Mohd Zubir Matjafri; Khiruddin Abdullah; Azlan Abdul Aziz; Abdul Aziz Tajuddin; Mohd Firdaus Othman

2005-01-01

An optical model has been developed for measuring total suspended solids TSS concentrations in water. This approach is based on the characteristics of scattered light from the suspended particles in water samples. An optical sensor system (an active spectrometer) has been developed to correlate pollutant (total suspended solids TSS) concentration and the scattered radiation. Scattered light was measured in terms of the output voltage of the phototransistor of the sensor system. The developed algorithm was used to calculate and estimate the concentrations of the polluted water samples. The proposed algorithm was calibrated using the observed readings. The results display a strong correlation between the radiation values and the total suspended solids concentrations. The proposed system yields a high degree of accuracy with the correlation coefficient (R) of 0.99 and the root mean square error (RMS) of 63.57 mg/l. (Author)

Research on physical and chemical parameters of coolant in Light-Water Reactors

Energy Technology Data Exchange (ETDEWEB)

Reis, Isabela C.; Mesquita, Amir Z., E-mail: icr@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEM-MG), Belo Horizonte, MG (Brazil)

2015-07-01

The coolant radiochemical monitoring of light-water reactors, both power reactor as research reactors is one most important tasks of the system safe operation. The last years have increased the interest in the coolant chemical studying to optimize the process, to minimize the corrosion, to ensure the primary system materials integrity, and to reduce the workers exposure radiation. This paper has the objective to present the development project in Nuclear Technology Development Center (CDTN), which aims to simulate the primary water physical-chemical parameters of light-water-reactors (LWR). Among these parameters may be cited: the temperature, the pressure, the pH, the electric conductivity, and the boron concentration. It is also being studied the adverse effects that these parameters can result in the reactor integrity. The project also aims the mounting of a system to control and monitoring of temperature, electric conductivity, and pH of water in the Installation of Test in Accident Conditions (ITCA), located in the Thermal-Hydraulic Laboratory at CDTN. This facility was widely used in the years 80/90 for commissioning of several components that were installed in Angra 2 containment. In the test, the coolant must reproduce the physical and chemical conditions of the primary. It is therefore fundamental knowledge of the main control parameters of the primary cooling water from PWR reactors. Therefore, this work is contributing, with the knowledge and the reproduction with larger faithfulness of the reactors coolant in the experimental circuits. (author)

The economics of the fuel cycle (light water reactors)

International Nuclear Information System (INIS)

Lepine, J.

1979-01-01

The economical characteristics of the fuel cycle (of light water reactors) as well as the definition and calculation method for the average updated cost of the kWh are recalled. The evolution followed by the unit prices of the different operations of the cycle, their total cost and the part taken by this cost in the overall cost of nuclear kWh are described. The effects on the cost of fuel of certain hypotheses, operating requirements and additional cost factors are considered [fr

Light water reactors fuel assembly mechanical design and evaluation

International Nuclear Information System (INIS)

Anon.

1981-01-01

This standard establishes a procedure for performing an evaluation of the mechanical design of fuel assemblies for light water-cooled commercial power reactors. It does not address the various aspects of neutronic or thermalhydraulic performance except where these factors impose loads or constraints on the mechanical design of the fuel assemblies. This standard also includes a set of specific requirements for design, various potential performance problems and criteria aimed specifically at averting them. This standard replaces ANSI/ANS-57.5-1978

Light Modulation and Water Splitting Enhancement Using a Composite Porous GaN Structure.

Science.gov (United States)

Yang, Chao; Xi, Xin; Yu, Zhiguo; Cao, Haicheng; Li, Jing; Lin, Shan; Ma, Zhanhong; Zhao, Lixia

2018-02-14

On the basis of the laterally porous GaN, we designed and fabricated a composite porous GaN structure with both well-ordered lateral and vertical holes. Compared to the plane GaN, the composite porous GaN structure with the combination of the vertical holes can help to reduce UV reflectance and increase the saturation photocurrent during water splitting by a factor of ∼4.5. Furthermore, we investigated the underlying mechanism for the enhancement of the water splitting performance using a finite-difference time-domain method. The results show that the well-ordered vertical holes can not only help to open the embedded pore channels to the electrolyte at both sides and reduce the migration distance of the gas bubbles during the water splitting reactions but also help to modulate the light field. Using this composite porous GaN structure, most of the incident light can be modulated and trapped into the nanoholes, and thus the electric fields localized in the lateral pores can increase dramatically as a result of the strong optical coupling. Our findings pave a new way to develop GaN photoelectrodes for highly efficient solar water splitting.

Photocatalytic Hydrogen or Oxygen Evolution from Water over S- or N-Doped TiO2 under Visible Light

Directory of Open Access Journals (Sweden)

Kazumoto Nishijima

2008-01-01

Full Text Available S- or N-doping of TiO2 powder having an anatase or rutile phase extended the photocatalytic activity for water oxidation and reduction under UV light and visible light irradiation. For the reduction of water, anatase-doped TiO2 showed higher level of activity than that of doped TiO2 having a rutile phase using ethanol as an electron donor. Furthermore, the activity level of S-doped TiO2 for hydrogen evolution was higher than that of N-doped TiO2 photocatalysts under visible light. Photocatalytic oxidation of water on doped TiO2 having a rutile phase proceeded with fairly high efficiency when Fe3+ ions were used as electron acceptors compared to that on doped TiO2 having an anatase phase. In addition, water splitting under visible light irradiation was achieved by construction of a Z-scheme photocatalysis system employing the doped TiO2 having anatase and rutile phases for H2 and O2 evolution and the I−/IO3− redox couple as an electron relay.

Changes in olive oil volatile organic compounds induced by water status and light environment in canopies of Olea europaea L. trees.

Science.gov (United States)

Benelli, Giovanni; Caruso, Giovanni; Giunti, Giulia; Cuzzola, Angela; Saba, Alessandro; Raffaelli, Andrea; Gucci, Riccardo

2015-09-01

Light and water are major factors in fruit development and quality. In this study, the effect of water and light in Olea europaea trees on volatile organic compounds (VOCs) in olive oil was studied over 2 years. Mature fruits were harvested from three zones of the canopy with different light exposure (64%, 42% and 30% of incident light) of trees subjected to full, deficit or complementary irrigation. VOCs were determined by SPME GC-MS and analysed by principal component analysis followed by discriminant analysis to partition treatment effects. Fruit fresh weight and mesocarp oil content decreased in zones where intercepted light was less. Low light levels significantly slowed down fruit maturation, whereas conditions of water deficit accelerated the maturation process. The presence of cyclosativene and α-muurulene was associated with water deficit, nonanal, valencene with full irrigation; α-muurulene, (E)-2-hexanal were related to low light conditions, while trans-β-ocimene, α-copaene, (Z)-2-penten-1-ol, hexanal and nonanal to well exposed zones. The year strongly affected the VOC profile of olive oil. This is the first report on qualitative changes in VOCs induced by light environment and/or water status. This information is valuable to better understand the role of environmental factors on the sensory quality of virgin olive oil. © 2014 Society of Chemical Industry.

Evolutionary/advanced light water reactor data report

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-02-09

The US DOE Office of Fissile Material Disposition is examining options for placing fissile materials that were produced for fabrication of weapons, and now are deemed to be surplus, into a condition that is substantially irreversible and makes its use in weapons inherently more difficult. The principal fissile materials subject to this disposition activity are plutonium and uranium containing substantial fractions of plutonium-239 uranium-235. The data in this report, prepared as technical input to the fissile material disposition Programmatic Environmental Impact Statement (PEIS) deal only with the disposition of plutonium that contains well over 80% plutonium-239. In fact, the data were developed on the basis of weapon-grade plutonium which contains, typically, 93.6% plutonium-239 and 5.9% plutonium-240 as the principal isotopes. One of the options for disposition of weapon-grade plutonium being considered is the power reactor alternative. Plutonium would be fabricated into mixed oxide (MOX) fuel and fissioned (``burned``) in a reactor to produce electric power. The MOX fuel will contain dioxides of uranium and plutonium with less than 7% weapon-grade plutonium and uranium that has about 0.2% uranium-235. The disposition mission could, for example, be carried out in existing power reactors, of which there are over 100 in the United States. Alternatively, new LWRs could be constructed especially for disposition of plutonium. These would be of the latest US design(s) incorporating numerous design simplifications and safety enhancements. These ``evolutionary`` or ``advanced`` designs would offer not only technological advances, but also flexibility in siting and the option of either government or private (e.g., utility) ownership. The new reactor designs can accommodate somewhat higher plutonium throughputs. This data report deals solely with the ``evolutionary`` LWR alternative.

Evolutionary/advanced light water reactor data report

International Nuclear Information System (INIS)

1996-01-01

The US DOE Office of Fissile Material Disposition is examining options for placing fissile materials that were produced for fabrication of weapons, and now are deemed to be surplus, into a condition that is substantially irreversible and makes its use in weapons inherently more difficult. The principal fissile materials subject to this disposition activity are plutonium and uranium containing substantial fractions of plutonium-239 uranium-235. The data in this report, prepared as technical input to the fissile material disposition Programmatic Environmental Impact Statement (PEIS) deal only with the disposition of plutonium that contains well over 80% plutonium-239. In fact, the data were developed on the basis of weapon-grade plutonium which contains, typically, 93.6% plutonium-239 and 5.9% plutonium-240 as the principal isotopes. One of the options for disposition of weapon-grade plutonium being considered is the power reactor alternative. Plutonium would be fabricated into mixed oxide (MOX) fuel and fissioned (''burned'') in a reactor to produce electric power. The MOX fuel will contain dioxides of uranium and plutonium with less than 7% weapon-grade plutonium and uranium that has about 0.2% uranium-235. The disposition mission could, for example, be carried out in existing power reactors, of which there are over 100 in the United States. Alternatively, new LWRs could be constructed especially for disposition of plutonium. These would be of the latest US design(s) incorporating numerous design simplifications and safety enhancements. These ''evolutionary'' or ''advanced'' designs would offer not only technological advances, but also flexibility in siting and the option of either government or private (e.g., utility) ownership. The new reactor designs can accommodate somewhat higher plutonium throughputs. This data report deals solely with the ''evolutionary'' LWR alternative

Numerical simulation of multi-phase phenomena in IVR related processes

Energy Technology Data Exchange (ETDEWEB)

Cheng, Xu [Karlsruher Institut fuer Technologie (KIT), Karlsruhe (Germany). Bereich Innovative Reaktorsysteme; Shanghai Jiao Tong Univ. (China). School of Nuclear Science and Engineering

2016-05-15

IVR (in-vessel retention) is one of the severe accident mitigation measures, which is widely applied in the advanced light water reactors (LWRs) such as KERENA of AREVA, AP1000 of Westinghouse and CAP1400 of SNPTC, and attracts extensive interests of the German and Chinese nuclear scientists. The ultimate target of IVR is to keep the core melt inside the reactor pressure vessel (RPV) and to provide cooling capability via water flowing outside the RPV, the so called external reactor vessel cooling (ERVC). This paper summarizes some activities ongoing in both KIT and SJTU (Shanghai Jiao Tong University) with a few results examples.

Four-photon parametric light scattering of ultrashort laser pulses in water in case of weak self-phase modulation

International Nuclear Information System (INIS)

Babenko, V A; Sychev, Andrei A

2009-01-01

The hyper-Raman scattering (HRS) of light in water is detected reliably by the active spectroscopy method of coherent light scattering, in particular, by the method of four-photon parametric light scattering in a medium in which HRS is a 'signal' wave in the parametric process involving simultaneously two high-power laser photons and IR photons of an 'idler' wave. Hyper-Raman scattering by libration vibrations of water molecules, which virtually cannot be detected by conventional methods of Raman scattering, was observed. (nonlinear optical phenomena)

Nondestructive Examination (NDE) Reliability for Inservice Inspection of Light Water Reactors

International Nuclear Information System (INIS)

Doctor, S.R.; Diaz, A.A.; Friley, J.R.; Good, M.S.; Greenwood, M.S.; Heasler, P.G.; Hockey, R.L.; Kurtz, R.J.; Simonen, F.A.; Spanner, J.C.; Taylor, T.T.; Vo, T.V.

1992-07-01

The Evaluation and Improvement of NDE Reliability for Inservice Inspection of Light Water Reactors (NDE Reliability) Program at the Pacific Northwest Laboratory was established by the Nuclear Regulatory Commission to determine the reliability of current inservice inspection (ISI) techniques and to develop recommendations that will ensure a suitably high inspection reliability. The objectives of this program include determining the reliability of ISI performed on the primary systems of commercial light-water reactors (LWR's); using probabilistic fracture mechanics analysis to determine the impact of NDE unreliability on system safety; and evaluating reliability improvements that can be achieved with improved and advanced technology. A final objective is to formulate recommended revisions to the Regulatory and ASME Code requirements, based on material properties, service conditions, and NDE uncertainties. The program scope is limited to ISI of the primary systems including the piping, vessel, and other components inspected in accordance with Section XI of the ASME Code. This is a progress report covering the programmatic work from April 1991 through September 1991

Water-light interaction: A novel pathway for multi hallmark therapy in cancer

Directory of Open Access Journals (Sweden)

Luis Santana-Blank

2014-02-01

Full Text Available Laser photobiomodulation (LPBM has been proposed as a multi-target (multi-hallmark therapy for cancer and other complex diseases based on an approach that aims to substitute and/or complement metabolic energy pathways through oxygen-dependent (e.g., cytochrome c oxidase (CcO and/or oxygen-independent (e.g., light-water interactions (e.g., F0-F1 motors mechanisms with critical signaling pathways in primarily aqueous media. Cellular and molecular bases for water-mediated, long-range, energy supplementation aimed at inducing and modulating physiologically reparative processes, including apoptosis, have been previously presented through a mechanism termed Photo Infrared Pulsed Biomodulation (PIPBM. Water’s role as an oscillator in LPBM has also been documented. These ideas were recently complemented by integrating the role of the quasi-crystalline exclusion zone (EZ described by Pollack as the fourth phase of water. This is retrospective analysis of experimental and clinical data using an infrared pulsed laser device (IPLD. It found photo-induced effects over the water dynamics of burned rat tissue monitored by 1H-NMR transverse relaxation times (1/T2, indicating significantly greater structuring of water. In addition, a microdensitometry study of T2 weighted tumor heterogeneities from a phase I clinical trial of the IPLD in patients with advanced neoplasias and an algorithm for tumor characterization indicated significantly increased structuring of water, possibly proving a photobiomodulation effect over the EZ associated with histologically-confirmed selective photo-induced tumor cell death. To the best of our knowledge, this is the first clinical demonstration of light-induced effects over the EZ. It supports our premise that LPBM can increase potential energy in the EZ, which then acts as a rechargeable electrolytic bio-battery for the external selective supplementation of the energy demand required for cellular work, signaling pathways and

Energy Research Advisory Board, Civilian Nuclear Power Panel: Subpanel 1 report, Light water reactor utilization and improvement: Volume 2

International Nuclear Information System (INIS)

1986-10-01

The Secretary of Energy requested that the Office of Nuclear Energy prepare a strategic national plan that outlines the Department's role in the future development of civilian nuclear power and that the Energy Research Advisory Board establish an ad hoc panel to review and comment on this plan. The Energy Research Advisory Board formed a panel for this review and three subpanels were formed. One subpanel was formed to address the institutional issues surrounding nuclear power, one on research and development for advanced nuclear power plants and a third subpanel on light water reactor utilization and improvement. The subpanel on light water reactors held two meetings at which representatives of the DOE, the NRC, EPRI, industry and academic groups made presentations. This is the report of the subpanel on light water reactor utilization and improvement. This report presents the subpanel's assessment of initiatives which the Department of Energy should undertake in the national interest, to develop and support light water reactor technologies

Results of the mid-term assessment of the 'High Performance Light Water Reactor Phase 2' project

International Nuclear Information System (INIS)

Starflinger, J.; Schulenberg, T.; Marsault, P.

2009-01-01

The High Performance Light Water Reactor (HPLWR) is a Light Water Reactor (LWR) operating at supercritical pressure (p>22.1 MPa). In Europe, investigations on the HPLWR have been integrated into a joint research project, called High Performance Light Water Reactor Phase 2 (HPLWR Phase 2), which is co-funded by the European Commission. Within the second year of the project, the design of the reactor core, the pressure vessel and its internals have been analysed in detail by means of advanced codes and methods. The mechanical design has been assessed and shows that stresses inside components and possible deformations keep within acceptable limits. The neutronics and the flow inside the core have been investigated. The addition of a water layer in the reflector helps to flatten the radial power profile. The moderator flow path must be changed because of possible reverse flow in the gaps between the assemblies (downward flow). First calculations of transients showed an acceptable behaviour of the cladding temperatures. Material oxidation experiments were successfully performed. The auxiliary loop of the Supercritical Water Loop has been constructed. Heat transfer has been investigated numerically analysing heat transfer deterioration (HTD) and flow around fuel pins with wire wrap spacers. (author)

More recent developments for the ultrasonic testing of light water reactor pressure vessels

International Nuclear Information System (INIS)

Seiger, H.; Engl, G.

1976-01-01

The development of an ultrasonic testing method for the inspection from the outside of the areas close to the cladding of the spherical fields of holes of light water reactor pressure vessels is described

Effects of light and microbial activity on the degradation of two fluoroquinolone antibiotics in pond water and sediment.

Science.gov (United States)

Lin, Juo-Shan; Pan, Hung-Yu; Liu, Shiu-Mei; Lai, Hong-Thih

2010-07-01

Enrofloxacin (ENR) and ciprofloxacin (CIP) are two fluoroquinolone (FQ) antibiotics widely used to treat diseases of human beings and cultured animals. These two FQs are usually detected in the effluent of municipal sewage plants and related aquatic environments. The purpose of this study was to understand the fates of ENR and CIP in aquaculture pond water and a sediment slurry in a laboratory-scale experiment. Effects of light and microbial activity on the degradation of these two FQs were investigated. Results indicated that natural irradiation plays a major role in the degradation of ENR and CIP in pond water and the sediment slurry. The 50 % dissipation times (DT(50)) with non-sterile treatment were 0.01 and 18.4 d for ENR, and 0.04 and 17.3 d for CIP in the water and sediment slurry, respectively. On the other hand, the degradation of ENR and CIP under dark conditions was slow or even hindered, and all of their DT(50) values exceeded 100 d. These two FQs degraded faster in the sediment slurry than in pond water under dark conditions. Artificial ultraviolet (UV) and fluorescence light had similar effects on the degradation of ENR in the pond water and sediment slurry. Degradation of CIP was faster with UV than with fluorescence light treatment, while no such difference was found for ENR degradation. CIP was a degradation product of ENR under both light and dark conditions, and DT(50) values for both compounds were shorter in the presence of light. The phenomenon of biodegradation was observed during degradation of CIP in the sediment slurry under natural light.

CAB models for water: A new evaluation of the thermal neutron scattering laws for light and heavy water in ENDF-6 format

International Nuclear Information System (INIS)

Márquez Damián, J.I.; Granada, J.R.; Malaspina, D.C.

2014-01-01

Highlights: • We present a new evaluation of the thermal scattering laws for light and heavy water. • This evaluation is based on molecular and experimental data, with no free parameters. • Calculations with these libraries compare well with experimental values. • Libraries result in an improvement over existing ENDF scattering law files. - Abstract: In this work we present the CAB models for water: a set of new models for the evaluation of the thermal neutron scattering laws for light and heavy water in ENDF-6 format, using the LEAPR module of NJOY. These models are based on experimental structure data and frequency spectra computed from molecular dynamics simulations. The calculations show a significant improvement over ENDF/B-VI and ENDF/B-VII when compared with measurements of differential and integral scattering data

Neutron collar calibration for assay of LWR [light-water reactor] fuel assemblies

International Nuclear Information System (INIS)

Menlove, H.O.; Pieper, J.E.

1987-03-01

The neutron-coincidence collar is used for the verification of the uranium content in light-water reactor fuel assemblies. An AmLi neutron source is used to give an active interrogation of the fuel assembly to measure the 235 U content, and the 238 U content is verified from a passive neutron-coincidence measurement. This report gives the collar calibration data of pressurized-water reactor and boiling-water reactor fuel assemblies. Calibration curves and correction factors are presented for neutron absorbers (burnable poisons) and different fuel assembly sizes. The data were collected at Exxon Nuclear, Franco-Belge de Fabrication de Combustibles, ASEA-Atom, and other nuclear fuel fabrication facilities

Occupational radiation exposure at Commercial Nuclear Power reactors 1983. Volume 5. Annual report

International Nuclear Information System (INIS)

Brooks, B.G.

1985-03-01

This report presents an updated compilation of occupational radiation exposure at commercial nuclear power reactors for the years 1969 through 1983. The summary based on information received from the 75 light-water-cooled reactors (LWRs) and one high temperature gas-cooled reactor (HTGR). The total number of personnel monitored at LWRs in 1983 was 136,700. The number of workers that received measurable doses during 1983 and 85,600 which is about 1000 more than that found in 1982. The total collective dose at LWRs for 1983 is estimated to be 56,500 man-rems (man-cSv), which is about 4000 more man-rems (man-cSv) than that reported in 1982. This resulted in the average annual dose for each worker who received a measurable dose increasing slightly to 0.66 rems (cSv), and the average collective dose per reactor increasing by about 50 man-rems (man-cSv), and the average collective dose per reactor increasing by about 50 man-rems (man-cSv) to a value of 753 man-rems (man-cSv). The collective dose per megawatt of electricity generated by each reactor also increased slightly to an average value of 1.7 man-rems (man-cSv) per megawatt-year. Health implications of these annual occupational doses are discussed

Basic policy of maintenance for the power conversion system of the gas turbine high temperature reactor 300 (GTHTR300)

International Nuclear Information System (INIS)

Kosugiyama, Shinichi; Takizuka, Takakazu; Kunitomi, Kazuhiko; Yan, Xing; Katanishi, Shoji; Takada, Shoji

2003-01-01

Basic policy of maintenance was determined for major equipment in the power conversion system of the Gas Turbine High Temperature Reactor 300 (GTHTR300). It was developed based on the current maintenance practice in Light Water Reactors (LWRs), High Temperature Engineering Test Reactor (HTTR) and conventional combined cycle power plants while taking into account of unique design features of GTHTR300. First, potential degradation phenomena in operations were identified and corresponding maintenance approaches were proposed for the equipment. Such degradations encountered typically in LWRs as corrosion, erosion and stress corrosion cracking are unlikely to occur since the working fluid of GTHTR300 is inert helium. Main causes of the degradations are high operating temperature and pressure. The gas turbine, compressor, generator, control valves undergo opening and dismantling maintenance in a suitable time interval. The power conversion vessel, heat exchanger vessel, primary system piping and heat exchanging tubes of precooler are subjected to in-service inspections similar to those done in LWRs. As turbine blades represent the severest material degradation because of their high-temperature and high-stress operating conditions, a lifetime management scheme was suggested for them. The longest interval of open-casing maintenance of the gas turbine is estimated to be six to seven years from technical point of view. Present study is entrusted from the Ministry of Education, Culture, Sports, Science and Technology of Japan. (author)

Energetics of semi-catalyzed-deuterium, light-water-moderated, fusion-fission toroidal reactors

International Nuclear Information System (INIS)

Jassby, D.L.; Towner, H.H.; Greenspan, E.; Schneider, A.; Misolovin, A.; Gilai, D.

1978-07-01

The semi-catalyzed-deuterium Light-Water Hybrid Reactor (LWHR) comprises a lithium-free light-water-moderated blanket with U 3 Si fuel driven by a deuterium-based fusion-neutron source, with complete burn-up of the tritium but almost no burn-up of the helium-3 reaction product. A one-dimensional model for a neutral-beam-driven tokamak plasma is used to determine the operating modes under which the fusion energy multiplication Q/sub p/ can be equal to or greater than 0.5. Thermonuclear, beam-target, and energetic-ion reactions are taken into account. The most feasible operating conditions for Q/sub p/ approximately 0.5 are tau/sub E/ = 2 to 4 x 10 14 cm -3 s, = 10 to 20 keV, and E/sub beam/ = 500 to 1000 keV, with approximately 40% of the fusion energy produced by beam-target reactions. Illustrative parameters of LWHRs are compared with those of an ignited D-T reactor

Transmutation of waste actinides in light water reactors

International Nuclear Information System (INIS)

Gorrell, T.C.

1979-04-01

Actinide recycle and transmutation calculations were made for three irradiation options of a light water reactor (LWR). The cases considered were: all actinides recycled in regular uranium fuel assemblies; transuranic actinides recycled in separate MOX assemblies with 235 U enrichment of uranium; and transuranic actinides recycled in separate MOX assemblies with plutonium enrichment of natural uranium. When all actinides were recycled in a uniform lattice, the transuranic inventory after ten recycles was 38% of the inventory accumulated without recycle. When the transuranics from two regular uranium assemblies were combined with those recycled from a MOX assembly, the transuranic inventory was reduced 50% after five recycles

Outline of design, manufacturing and installation experience of pressure vessel structure for the prototype heavy water moderated boiling light water cooled reactor 'FUGEN'

International Nuclear Information System (INIS)

Shibato, Eizo; Oguchi, Isao; Kishi, Toshikazu; Kitagawa, Yuji

1977-01-01

After component installation completed in June 1977 and various functional tests to be conducted later, the prototype heavy water moderated, boiling light water cooled reactor ''FUGEN'' is scheduled to reach first criticality in March 1978. Since the pressure vessel of ''FUGEN'' is completely different from that of a light water reactor in structure and materials, through research and development work was carried out prior to fabrication and construction. Based on these studies, installation of the actual pressure vessel was completed. Functional tests are now under way. This article describes examples in which our research and development results are reflected on design, manufacture, and installation of the pressure vessel. Also it introduces noteworthy achievements relevant to production techniques in manufacture and installation. (auth.)